Safety of Nuclear Power Plants: Design

PAGE

SSR-2/1 (Rev. 1) Safety of Nuclear Power Plants: Design 1. Introduction Background

1.1. The present publication supersedes the Safety Requirements publication Safety of Nuclear Power Plants: Design¹1, which was issued in 2012 as IAEA Safety Standards Series No. SSR-2/1. Account has been taken of the Fundamental Safety Principles [1], published in 2006. Requirements for nuclear safety are intended to ensure “the highest standards of safety that can reasonably be achieved” for the protection of workers, the public and the environment from harmful effects of ionizing radiation that could arise from nuclear power plants and other nuclear facilities [1]. It is recognized that technology and scientific knowledge advance, and that nuclear safety and the adequacy of protection against radiationrisks need to be considered in the context of the present state of knowledge. Safety requirements will change over time; this Safety Requirements publication reflects the present consensus.

1.2. The designs of many existing nuclear power plants, as well as the designs for new nuclear power plants, have been enhanced to include additional measures to mitigate the consequences of complex accident sequences involving multiple failures and of severe accidents. Complementary systems and equipment with new capabilities have been backfitted to many existing nuclear power plants to aid in the prevention of severe accidents and the mitigation of their consequences. Guidance on the mitigation of the consequences of severe accidents has been provided at most existing nuclear power plants. The design of new nuclear power plants now explicitly includes the consideration of severe accidentscenarios and strategies for their management. Requirements related to the State system of accounting for, and control of, nuclear material and security related requirements are also taken into account in the design of nuclear power plants. Integration of safety measures and security measures will help to ensure that neither compromise the other.

1.3. It might not be practicable to apply all the requirements of this Safety Requirements publication to nuclear power plants that are already in operation or under construction. In addition, it might not be feasible to modify designs that have already been approved by regulatory bodies. For the safety analysis of such designs, it is expected that a comparison will be made with the current standards, for example as part of the periodic safety review for the plant, to determine whether the safe operation of the plant could be further enhanced by means of reasonably practicable safety improvements.

All the General Safety Requirements, GSR Part 1 (Rev. 1), GSR Part 2, GSR Part 3, GSR Part 4 (Rev. 1), GSR Part 5, GSR Part 6 and GSR Part 7 are also applicable, as well as the Specific Safety Requirement SSR-1 on Site Evaluation for Nuclear Installations and SSR-2/2 (Rev. 1) on Safety of Nuclear Power Plants: Commissioning and Operation.

Here below, in front of specific requirements, you will also find information on the relevant Safety Guides for these specific topical areas.

You may also wish to read:

TECDOC No. 1936 on Applicability of Design Safety Requirements to Small Modular Reactor Technology Intended for Near Term Deployment

TECDOC No. 1791 on Considerations on the Application of the IAEA Safety Requirements for the Design of Nuclear Power Plants

IAEA Services Series 41 Technical Service Review (TSR) Service Guidelines, Vienna 2019

IAEA-TECDOC-2010 Approach and Methodology for the Development of Regulatory Safety Requirements for the Design of Advanced Nuclear Power Reactors, Case Study on Small Modular Reactors
Objective
1.4. This publication establishes designrequirements for the structures, systems and components of a nuclear power plant, as well as for procedures and organizational processes important to safety that are required to be met for safe operation and for preventing events that could compromise safety, or for mitigating the consequences of such events, were they to occur.

1.5. This publication is intended for use by organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning for nuclear power plants, in analysis, verification and review, and in the provision of technical support, as well as by regulatory bodies.
Scope
1.6. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). This publication may also be applied, with judgement, to other reactor types, to determine the requirements that have to be considered in developing the design.

1.7. This publication does not address:

Requirements that are specifically covered in other IAEA Safety Requirements publications (e.g. IAEA Safety Standards Series No. GSR Part 4 (Rev. 1), Safety Assessment for Facilities and Activities [2]);

Matters relating to nuclear security or to the State system of accounting for, and control of, nuclear material;

Conventional industrial safety that under no circumstances could affect the safety of the nuclear power plant;

Non-radiological impacts arising from the operation of nuclear power plants.

1.8. Terms in this publication are to be understood as defined and explained in the IAEA Safety Glossary [3], unless otherwise stated here (see Definitions).
Structure
1.9. This Safety Requirements publication follows the relationship between the safety objective and safety principles, and between requirements for nuclearsafety functions and design criteria for safety. Section 2 elaborates on the safety objective, safety principles and concepts that form the basis for deriving the safety functionrequirements that must be met for the nuclear power plant, as well as the safetydesign criteria. Sections 3–6 establish numbered overarching requirements (shown in bold type), with additional requirements as appropriate in the paragraphs that follow them. Section 3 establishes the general requirements to be satisfied by the design organization in the management of safety in the design process. Section 4 establishes: requirements for principal technical design criteria for safety, including requirements for the fundamental safety functions, the application of defence in depth and provision for construction; requirements for interfaces of safety with nuclear security and with the State system of accounting for, and control of, nuclear material; and requirements for ensuring that radiationrisks arising from the plant are maintained as low as reasonably achievable. Section 5 establishes requirements for general plant design that supplement the requirements for principal technical design criteria to ensure that safety objectives are met and the safety principles are applied. The requirements for general plant design apply to all items (i.e. structures, systems and components) important to safety. Section 6 establishes requirements for the design of specific plant systems such as the reactor core, reactor coolant systems, containment system, and instrumentation and control systems.
2. Applying the safety principles and concepts
2.1. The Fundamental Safety Principles [1] establish one fundamental safety objective and ten safety principles that provide the basis for requirements and measures for the protection of people and the environment against radiation risks and for the safety of facilities and activities that give rise to radiation risks.

2.2. This fundamental safety objective has to be achieved, and the ten safety principles have to be applied, without unduly limiting the operation of facilities or the conduct of activities that give rise to radiation risks. To ensure that nuclear power plants are operated and activities are conducted so as to achieve the highest standards of safety that can reasonably be achieved, measures have to be taken to achieve the following (see para. 2.1 of the Fundamental Safety Principles [1]):

To control the radiation exposure of people and radioactivereleases to the environment in operational states;

To restrict the likelihood of events that might lead to a loss of control over a nuclear reactor core, nuclear chain reaction, radioactive source, spent nuclear fuel, radioactive waste or any other source of radiation at a nuclear power plant;

To mitigate the consequences of such events, were they to occur.

2.3. The fundamental safety objective applies for all stages in the lifetime of a nuclear power plant, including planning, siting, design, manufacture, construction, commissioning and operation, as well as decommissioning. This includes the associated transport of radioactive material and the management of spent nuclear fuel and radioactive waste (see para. 2.2 of the Fundamental Safety Principles [1]).

2.4. Paragraph 2.3 of the Fundamental Safety Principles [1] states that:

“Ten safety principles have been formulated, on the basis of which safety requirements are developed and safety measures are to be implemented in order to achieve the fundamental safety objective. The safety principles form a set that is applicable in its entirety; although in practice different principles may be more or less important in relation to particular circumstances, the appropriate application of all relevant principles is required.”

2.5. This Safety Requirements publication establishes requirements that apply those safety principles, which are particularly important in the design of nuclear power plants.
Radiation protection in design
2.6. In order to satisfy the safety principles, it is required to ensure that for all operational states of a nuclear power plant and for any associated activities, doses from exposure to radiation within the installation or exposure due to any planned radioactiverelease from the installation are kept below the dose limits and kept as low as reasonably achievable. In addition, it is required to take measures for mitigating the radiological consequences of any accidents, if they were to occur.

2.7. To apply the safety principles, it is also required that nuclear power plants be designed and operated so as to keep all sources of radiation under strict technical and administrative control. However, this principle does not preclude limitedexposures or the release of authorized amounts of radioactive substances to the environment from nuclear power plants in operational states. Such exposures and radioactivereleases are required to be strictly controlled and to be kept as low as reasonably achievable, in compliance with regulatory and operational limits as well as radiation protectionrequirements [4].
Safety in design
2.8. To achieve the highest level of safety that can reasonably be achieved in the design of a nuclear power plant, measures are required to be taken to do the following, consistent with national acceptance criteria and safety objectives [1]:

To prevent accidents with harmful consequences resulting from a loss of control over the reactor core or over other sources of radiation, and to mitigate the consequences of any accidents that do occur;

To ensure that for all accidents taken into account in the design of the installation, any radiological consequences would be below the relevant limits and would be kept as low as reasonably achievable;

To ensure that the likelihood of occurrence of an accident with serious radiological consequences is extremely low and that the radiological consequences of such an accident would be mitigated to the fullest extent practicable.

2.9. To demonstrate that the fundamental safety objective [1] is achieved in the design of a nuclear power plant, a comprehensive safety assessment [2] of the design is required to be carried out. Its objective is to identify all possible sources of radiation and to evaluate possible doses that could be received by workers at the installation and by members of the public, as well as possible effects on the environment, as a result of operation of the plant. The safety assessment is required in order to examine: (i) normal operation of the plant; (ii) the performance of the plant in anticipated operational occurrences; and (iii) accident conditions. On the basis of this analysis, the capability of the design to withstand postulated initiating events and accidents can be established, the effectiveness of the items important to safety can be demonstrated and the inputs (prerequisites) for emergency planning can be established.

2.10. Measures are required to be taken to controlexposure for all operational states at levels that are as low as reasonably achievable and to minimize the likelihood of an accident that could lead to the loss of control over a source of radiation. Nevertheless, there will remain a possibility that an accident could happen. Measures are required to be taken to ensure that the radiological consequences of an accident would be mitigated. Such measures include the provision of safety features and safety systems, the establishment of accident managementprocedures by the operating organization and, possibly, the establishment of off-siteprotective actions by the appropriate authorities, supported as necessary by the operating organization, to mitigate exposures if an accident occurs.

2.11. The design for safety of a nuclear power plant applies the safety principle that practical measures must be taken to mitigate the consequences for human life and health and for the environment of nuclear or radiationaccidents (Principle 8 of the Fundamental Safety Principles [1]. Plant event sequences that could result in high radiationdoses or in a large radioactive release have to be ‘practically eliminated’²2and plant event sequences with a significant frequency of occurrence have to have no, or only minor, potential radiological consequences. An essential objective is that the necessity for off-siteprotective actions to mitigate radiological consequences be limited or even eliminated in technical terms, although such measures might still be required by the responsible authorities.
The concept of defence in depth
2.12. The primary means of preventing accidents in a nuclear power plant and mitigating the consequences of accidents if they do occur is the application of the concept of defence in depth [1, 5, 6]. This concept is applied to all safety related activities, whether organizational, behavioural or design related, and whether in full power, low power or various shutdown states. This is to ensure that all safety related activities are subject to independent layers of provisions so that if a failure were to occur, it would be detected and compensated for or corrected by appropriate measures. Application of the concept of defence in depth throughout design and operation provides protection against anticipated operational occurrences and accidents, including those resulting from equipment failure or human induced events within the plant, and against consequences of events that originate outside the plant.

2.13. Paragraph 3.31 of the Fundamentals Safety Principles [1] states that:

“Defence in depth is implemented primarily through the combination of a number of consecutive and independent levels of protection that would have to fail before harmful effects could be caused to people or to the environment. If one level of protection or barrier were to fail, the subsequent level or barrier would be available. The independent effectiveness of the different levels of defence is a necessary element of defence in depth”.

There are five levels of defence:

The purpose of the first level of defence is to prevent deviations from normal operation and the failure of items important to safety. This leads to requirements that the plant be soundly and conservatively sited, designed, constructed, maintained and operated in accordance with quality management and appropriate and proven engineering practices. To meet these objectives, careful attention is paid to the selection of appropriate design codes and materials, and to the quality control of the manufacture of components and construction of the plant, as well as to its commissioning. Design options that reduce the potential for internal hazards contribute to the prevention of accidents at this level of defence. Attention is paid to the processes and procedures involved in design, manufacture, construction and in-service inspection, maintenance and testing, to the ease of these activities, and to the way the plant is operated and to how operating experience is utilized. This process is supported by a detailed analysis that determines the requirements for operation and maintenance of the plant and requirements for quality management for operational and maintenancepractices.

The purpose of the second level of defence is to detect and controldeviations from normal operational states in order to prevent anticipated operational occurrences at the plant from escalating to accident conditions. This is in recognition of the fact that postulated initiating events are likely to occur over the operating lifetime of a nuclear power plant, despite the care taken to prevent them. This second level of defence necessitates the provision of specific systems and features in the design, the confirmation of their effectiveness through safety analysis, and the establishment of operating procedures to prevent such initiating events, or otherwise to minimize their consequences, and to return the plant to a safe state.

For the third level of defence, it is assumed that, although very unlikely, the escalation of certain anticipated operational occurrences or postulated initiating events might not be controlled at a preceding level and that an accident could develop. In the design of the plant, such accidents are postulated to occur. This leads to the requirement that inherent and/or engineered safety features, safety systems and procedures be capable of preventing damage to the reactor core or preventing radioactivereleases requiring off-siteprotective actions and returning the plant to a safe state.

The purpose of the fourth level of defence is to mitigate the consequences of accidents that result from failure of the third level of defence in depth. This is achieved by preventing the progression of such accident and mitigating the consequences of a severe accident. The safety objective in the case of a severe accident is that only protective actions that are limited in terms of lengths of times and areas of application would be necessary and that off-sitecontamination would be avoided or minimized. Event sequences that would lead to an early radioactive release or a large radioactive release³3 are required to be ‘practically eliminated’⁴4.

The purpose of the fifth and final level of defence is to mitigate the radiological consequences of radioactivereleases that could potentially result from accidents. This requires the provision of an adequately equipped emergency response facilities and emergency plans and emergency procedures for on-site and off-siteemergency response.

2.14. A relevant aspect of the implementation of defence in depth for a nuclear power plant is the provision in the design of a series of physical barriers, as well as a combination of active, passive and inherent safety features that contribute to the effectiveness of the physical barriers in confining radioactive material at specified locations. The number of barriers that will be necessary will depend upon the initial source term in terms of the amount and isotopic composition of radionuclides, the effectiveness of the individual barriers, the possible internal and external hazards, and the potential consequences of failures.

You may also wish to read:

INSAG-28, Application of the Principle of Defence in Depth in Nuclear Safety to Small Modular Reactors Addendum to INSAG-10
Maintaining the integrity of design of the plant throughout the lifetime of the plant
2.15. The design, construction and commissioning of a nuclear power plant might be shared between a number of organizations: the architect–engineer, the vendor of the reactor and its supporting systems, the suppliers of major components, the designers of electrical systems, and the suppliers of other systems that are important to the safety of the plant.

2.16. The prime responsibility for safety rests with the person or organization responsible for facilities and activities that give rise to radiation risks (i.e. the operating organization) [1]. In 2003, the International Nuclear Safety Advisory Group suggested that the operating organization could set up a formal process to maintain the integrity of design of the plant throughout the lifetime of the plant (i.e. during the operating lifetime and into the decommissioning stage) [7]. A formally designated entity within the operating organization would take responsibility for this process.

2.17. In practice, the design of a nuclear power plant is complete only when the full plant specification (including site details) is produced for its procurement and licensing. Reference [7] emphasizes the need for a formally designated entity that has overall responsibility for the design process and is responsible for approving design changes and for ensuring that the requisite knowledge is maintained. Reference [7]also introduces the concept of ‘responsible designers’, to whom this formally designated entity could assign specific responsibilities for the design of parts of the plant. Prior to an application for authorization of a plant, the responsibility for the design will rest with the design organization (e.g. the vendor). Once an application for authorization of a plant has been made, the prime responsibility for safety will lie with the applicant, although detailed knowledge of the design will rest with the responsible designers. This balance will change as the plant is put into operation, since much of this detailed knowledge, such as the knowledge embodied in the safety analysis report, design manuals and other design documentation, will be transferred to the operating organization. To facilitate this transfer of knowledge, the structure of the formally designated entity that has overall responsibility for the design process would be established at an early stage.

2.18. The management systemrequirements that are placed on this formally designated entity would also apply to the responsible designers. However, the overall responsibility for maintaining the integrity of design of the plant would rest with the formally designated entity, and hence, ultimately, with the operating organization.
3. Management of safety in design Requirement 1: Responsibilities in the management of safety in plant design
An applicant for a licence to construct and/or operate a nuclear power plant shall be responsible for ensuring that the design submitted to the regulatory body meets all applicable safety requirements.

3.1. All organizations, including the design organization,⁵5 engaged in activities important to the safety of the design of a nuclear power plant shall be responsible for ensuring that safety matters are given the highest priority.

See also GSR Part 2 on Leadership and Management for Safety, GS-G-3.1 on Application of the Management System for Facilities and Activities and GS-G-3.5 on The Management System for Nuclear Installations
Requirement 2: Management system for plant design
The design organization shall establish and implement a management system for ensuring that all safety requirements established for the design of the plant are considered and implemented in all phases of the design process and that they are met in the final design.

3.2. The management system⁶6 shall include provision for ensuring the quality of the design of each structure, system and component, as well as of the overall design of the nuclear power plant, at all times. This includes the means for identifying and correcting design deficiencies, for checking the adequacy of the design and for controllingdesign changes.

See also GSR Part 2 on Leadership and Management for Safety, GS-G-3.1 on Application of the Management System for Facilities and Activities and GS-G-3.5 on The Management System for Nuclear Installations

3.3. The design of the plant, including subsequent changes, modifications or safety improvements, shall be in accordance with established procedures that call on appropriate engineering codes and standards and shall incorporate relevant requirements and design bases. Interfaces shall be identified and controlled.

3.4. The adequacy of the plant design, including design tools and design inputs and outputs, shall be verified and validated by individuals or groups separate from those who originally performed the design work. Verification, validation and approval of the plant design shall be completed as soon as is practicable in the design and constructionprocesses, and in any case before operation of the plant is commenced.
Requirement 3: Safety of the plant design throughout the lifetime of the plant
The operating organization shall establish a formal system for ensuring the continuing safety of the plant design throughout the lifetime of the nuclear power plant.

3.5. The formal system for ensuring the continuing safety of the plant design shall include a formally designated entity responsible for the safety of the plant design within the operating organization’s management system. Tasks that are assigned to external organizations (referred to as responsible designers) for the design of specific parts of the plant shall be taken into account in the arrangements.

See also GSR Part 2 on Leadership and Management for Safety and GSR Part 4 (Rev. 1) on Safety Assessment for Facilities and Activities

3.6. The formally designated entity shall ensure that the plant design meets the acceptance criteria for safety, reliability and quality in accordance with relevant national and international codes and standards, laws and regulations. A series of tasks and functions shall be established and implemented to ensure the following:

That the plant design is fit for purpose and meets the requirement for the optimization of protection and safety by keeping radiation risksas low as reasonably achievable;

That the designverification, definition of engineering codes and standards and requirements, use of proven engineering practices, provision for feedback of information on construction and experience, approval of key engineering documents, conduct of safety assessments and maintaining a safety culture are included in the formal system for ensuring the continuing safety of the plant design;

That the knowledge of the design that is needed for safe operation, maintenance (including adequate intervals for testing) and modification of the plant is available, that this knowledge is maintained up to date by the operating organization, and that due account is taken of past operating experience and validated research findings;

That management of designrequirements and configurationcontrol are maintained;

That the necessary interfaces with responsible designers and suppliers engaged in design work are established and controlled;

That the necessary engineering expertise and scientific and technical knowledge are maintained within the operating organization;

That all design changes to the plant are reviewed, verified, documented and approved;

That adequate documentation is maintained to facilitate future decommissioning of the plant.

4. Principal technical requirements Requirement 4: Fundamental safety functions
Fulfilment of the following fundamental safety functions for a nuclear power plant shall be ensured for all plant states: (i) control of reactivity; (ii) removal of heat from the reactor and from the fuel store; and (iii) confinement of radioactive material, shielding against radiation and control of planned radioactivereleases, as well as limitation of accidental radioactivereleases.

4.1. A systematic approach shall be taken to identifying those items important to safety that are necessary to fulfil the fundamental safety functions and to identifying the inherent features that are contributing to fulfilling, or that are affecting, the fundamental safety functions for all plant states.

4.2. Means of monitoring the status of the plant shall be provided for ensuring that the required safety functions are fulfilled.
Requirement 5: Radiation protection in design
The design of a nuclear power plant shall be such as to ensure that radiationdoses to workers at the plant and to members of the public do not exceed the dose limits, that they are kept as low as reasonably achievable in operational states for the entire lifetime of the plant, and that they remain below acceptable limits and as low as reasonably achievable in, and following, accident conditions

4.3 The design shall be such as to ensure that plant states that could lead to high radiationdoses or a large radioactive release have been ‘practically eliminated’⁷7, and that there would be no, or only minor, potential radiological consequences for plant states with a significant likelihood of occurrence.

4.4. Acceptable limits for radiation protection⁸8 associated with the relevant categories of plant states shall be established, consistent with the regulatory requirements.

See also GSR Part 3Radiation Protection and Safety of Radiation Sources, International Basic Safety Standards and SSG-90 on Radiation Protection Aspects of Design for Nuclear Power Plants
Requirement 6: Design for a nuclear power plant
The design for a nuclear power plant shall ensure that the plant and items important to safety have the appropriate characteristics to ensure that safety functions can be performed with the necessary reliability, that the plant can be operated safely within the operational limits and conditions for the full duration of its design life and can be safely decommissioned, and that impacts on the environment are minimized.

4.5. The design for a nuclear power plant shall be such as to ensure that the safety requirements of the operating organization, the requirements of the regulatory body and the requirements of relevant legislation, as well as applicable national and international codes and standards, are all met, and that due account is taken of human capabilities and limitations and of factors that could influence human performance. Adequate information on the design shall be provided for ensuring the safe operation and maintenance of the plant, and to allow subsequent plant modifications to be made. Recommended practices shall be provided for incorporation into the administrative and operational procedures for the plant (i.e. the operational limits and conditions).

4.6. The design shall take due account of relevant available experience that has been gained in the design, construction and operation of other nuclear power plants, and of the results of relevant research programmes.

4.7. The design shall take due account of the results of deterministic safety analyses and probabilistic safety analyses, to ensure that due consideration is given to the prevention of accidents and to mitigation of the consequences of any accidents that do occur.

4.8. The design shall be such as to ensure that the generation of radioactive waste and discharges are kept to the minimum practicable in terms of both activity and volume, by means of appropriate design measures and operational and decommissioningpractices.
Requirement 7: Application of defence in depth
The design of a nuclear power plant shall incorporate defence in depth. The levels of defence in depth shall be independent as far as is practicable.

See also para. 2.12 to 2.14 above

4.9. The defence in depth concept shall be applied to provide several levels of defence that are aimed at preventing consequences of accidents that could lead to harmful effects on people and the environment, and ensuring that appropriate measures are taken for the protection of people and the environment and for the mitigation of consequences in the event that prevention fails.

4.10. The design shall take due account of the fact that the existence of multiple levels of defence is not a basis for continued operation in the absence of one level of defence. All levels of defence in depth shall be kept available at all times and any relaxations shall be justified for specific modes of operation.

4.11. The design:

Shall provide for multiple physical barriers to the release of radioactive material to the environment;

Shall be conservative, and the construction shall be of high quality, so as to provide assurance that failures and deviations from normal operation are minimized, that accidents are prevented as far as is practicable and that a small deviation in a plant parameter does not lead to a cliff edge effect;⁹9

Shall provide for the control of plant behaviour by means of inherent and engineered features, such that failures and deviations from normal operation requiring actuation of safety systems are minimized or excluded by design, to the extent possible;

Shall provide for supplementing the control of the plant by means of automatic actuation of safety systems, such that failures and deviations from normal operation that exceed the capability of control systems can be controlled with a high level of confidence, and the need for operator actions in the early phase of these failures or deviations from normal operation is minimized;

Shall provide for systems, structures and components and procedures to control the course of and, as far as practicable, to limit the consequences of failures and deviations from normal operation that exceed the capability of safety systems;

Shall provide multiple means for ensuring that each of the fundamental safety functions is performed, thereby ensuring the effectiveness of the barriers and mitigating the consequences of any failure or deviation from normal operation.

4.12. To ensure that the concept of defence in depth is maintained, the design shall prevent, as far as is practicable:

Challenges to the integrity of physical barriers;

Failure of one or more barriers;

Failure of a barrier as a consequence of the failure of another barrier;

The possibility of harmful consequences of errors in operation and maintenance.

4.13. The design shall be such as to ensure, as far as is practicable, that the first, or at most the second, level of defence is capable of preventing an escalation to accident conditions for all failures or deviations from normal operation that are likely to occur over the operating lifetime of the nuclear power plant.

4.13.A. The levels of defence in depth shall be independent as far as practicable to avoid the failure of one level reducing the effectiveness of other levels. In particular, safety features for design extension conditions (especially features for mitigating the consequences of accidents involving the melting of fuel) shall as far as is practicable be independent of safety systems.
Requirement 8: Interfaces of safety with security and safeguards
Safety measures, nuclear security measures and arrangements for the State system of accounting for, and control of, nuclear material for a nuclear power plant shall be designed and implemented in an integrated manner so that they do not compromise one another.

See also the Nuclear Security Series publications, also available in NSS-OUI
Requirement 9: Proven engineering practices
Items important to safety for a nuclear power plant shall be designed in accordance with the relevant national and international codes and standards.

4.14. Items important to safety for a nuclear power plant shall preferably be of a design that has previously been proven in equivalent applications, and if not, shall be items of high quality and of a technology that has been qualified and tested.

4.15. National and international codes and standards that are used as design rules for items important to safety shall be identified and evaluated to determine their applicability, adequacy and sufficiency, and shall be supplemented or modified as necessary to ensure that the quality of the design is commensurate with the associated safety function.

4.16. Where an unproven design or feature is introduced or where there is a departure from an established engineering practice, safety shall be demonstrated by means of appropriate supporting research programmes, performance tests with specific acceptance criteria or the examination of operating experience from other relevant applications. The new design or feature or new practice shall also be adequately tested to the extent practicable before being brought into service, and shall be monitored in service to verify that the behaviour of the plant is as expected.
Requirement 10: Safety assessment
Comprehensive deterministic safety assessments and probabilistic safety assessments shall be carried out throughout the design process for a nuclear power plant to ensure that all safety requirements on the design of the plant are met throughout all stages of the lifetime of the plant, and to confirm that the design, as delivered, meets requirements for manufacture and for construction, and as built, as operated and as modified.

4.17. The safety assessments¹⁰10 shall be commenced at an early point in the design process, with iterations between designactivities and confirmatory analytical activities, and shall increase in scope and level of detail as the design programme progresses.

See also GSR Part 4 (Rev. 1) on Safety Assessment for Facilities and Activities

4.18. The safety assessments shall be documented in a form that facilitates independent evaluation.
Requirement 11: Provision for construction
Items important to safety for a nuclear power plant shall be designed so that they can be manufactured, constructed, assembled, installed and erected in accordance with established processes that ensure the achievement of the design specifications and the required level of safety.

4.19. In the provision for construction and operation, due account shall be taken of relevant experience that has been gained in the construction of other similar plants and their associated structures, systems and components. Where best practices from other relevant industries are adopted, such practices shall be shown to be appropriate to the specific nuclear application.

See also SSG-38 on Construction for Nuclear Installations
Requirement 12: Features to facilitate radioactive waste management and decommissioning
Special consideration shall be given at the design stage of a nuclear power plant to the incorporation of features to facilitate radioactive waste management and the future decommissioning and dismantling of the plant.

4.20. In particular, the design shall take due account of:

The choice of materials, so that amounts of radioactive waste will be minimized to the extent practicable and decontamination will be facilitated;

The access capabilities and the means of handling that might be necessary;

The facilities necessary for the management (i.e. segregation, characterization, classification, pretreatment, treatment and conditioning) and storage of radioactive waste generated in operation and provision for managing the radioactive waste that will be generated in the decommissioning of the plant.

Se also GSR Part 5 on Predisposal Management of Radioactive Waste and GSR Part 6 on Decommissioning of Facilities

5. General plant design Design basis Requirement 13: Categories of plant states
Plant states shall be identified and shall be grouped into a limited number of categories primarily on the basis of their frequency of occurrence at the nuclear power plant.

5.1. Plant states shall typically cover:

Normal operation;

Anticipated operational occurrences, which are expected to occur over the operating lifetime of the plant;

Design basis accidents;

Design extension conditions, including accidents with core melting.

5.2. Criteria shall be assigned to each plant state, such that frequently occurring plant states shall have no, or only minor, radiological consequences and plant states that could give rise to serious consequences shall have a very low frequency of occurrence.
Requirement 14: Design basis for items important to safety
The design basis for items important to safety shall specify the necessary capability, reliability and functionality for the relevant operational states, for accident conditions and for conditions arising from internal and external hazards, to meet the specific acceptance criteria over the lifetime of the nuclear power plant.

5.3. The design basis for each item important to safety shall be systematically justified and documented. The documentation shall provide the necessary information for the operating organization to operate the plant safely.
Requirement 15: Design limits
A set of designlimits consistent with the key physical parameters for each item important to safety for the nuclear power plant shall be specified for all operational states and for accident conditions.

5.4. The designlimits shall be specified and shall be consistent with relevant national and international standards and codes, as well as with relevant regulatory requirements.
Requirement 16: Postulated initiating events
The design for the nuclear power plant shall apply a systematic approach to identifying a comprehensive set of postulated initiating events such that all foreseeable events with the potential for serious consequences and all foreseeable events with a significant frequency of occurrence are anticipated and are considered in the design.

5.5. The postulated initiating events shall be identified on the basis of engineering judgement and a combination of deterministic assessment and probabilistic assessment. A justification of the extent of usage of deterministic safety analysis and probabilistic safety analysis shall be provided to show that all foreseeable events have been considered.

5.6. The postulated initiating events shall include all foreseeable failures of structures, systems and components of the plant, as well as operating errors and possible failures arising from internal and external hazards, whether in full power, low power or shutdown states.

5.7. An analysis of the postulated initiating events for the plant shall be made to establish the preventive measures and protective measures that are necessary to ensure that the required safety functions will be performed.

5.8. The expected behaviour of the plant in any postulated initiating event shall be such that the following conditions can be achieved, in order of priority:

A postulated initiating event would produce no safety significant effects or would produce only a change towards safe plant conditions by means of inherent characteristics of the plant.

Following a postulated initiating event, the plant would be rendered safe by means of passive safety features or by the action of systems that are operating continuously in the state necessary to control the postulated initiating event.

Following a postulated initiating event, the plant would be rendered safe by the actuation of safety systems that need to be brought into operation in response to the postulated initiating event.

Following a postulated initiating event, the plant would be rendered safe by following specified procedures.

5.9. The postulated initiating events used for developing the performance requirements for the items important to safety in the overall safety assessment and the detailed analysis of the plant shall be grouped into a specified number of representative event sequences that identify bounding cases and that provide the basis for the design and the operational limits for items important to safety.

5.10. A technically supported justification shall be provided for exclusion from the design of any initiating event that is identified in accordance with the comprehensive set of postulated initiating events.

5.11. Where prompt and reliable action would be necessary in response to a postulated initiating event, provision shall be made in the design for automatic safety actions for the necessary actuation of safety systems, to prevent progression to more severe plant conditions.

5.12. Where prompt action in response to a postulated initiating event would not be necessary, it is permissible for reliance to be placed on the manual initiation of systems or on other operator actions. For such cases, the time interval between detection of the abnormal event or accident and the required action shall be sufficiently long, and adequate procedures (such as administrative, operational and emergency procedures) shall be specified to ensure the performance of such actions. An assessment shall be made of the potential for an operator to worsen an event sequence through erroneous operation of equipment or incorrect diagnosis of the necessary recovery process.

5.13. The operator actions that would be necessary to diagnose the state of the plant following a postulated initiating event and to put it into a stable long term shutdowncondition in a timely manner shall be facilitated by the provision of adequate instrumentation to monitor the status of the plant, and adequate controls for the manual operation of equipment.

5.14. The design shall specify the necessary provision of equipment and the procedures necessary to provide the means for keeping control over the plant and for mitigating any harmful consequences of a loss of control.

5.15. Any equipment that is necessary for actions to be taken in manual response and recovery processes shall be placed at the most suitable location to ensure its availability at the time of need and to allow safe access to it under the environmental conditions anticipated.
Requirement 17: Internal and external hazards
All foreseeable internal hazards and external hazards, including the potential for human induced events directly or indirectly to affect the safety of the nuclear power plant, shall be identified and their effects shall be evaluated. Hazards shall be considered in designing the layout of the plant and in determining the postulated initiating events and generated loadings for use in the design of relevant items important to safety for the plant.

5.15A. Items important to safety shall be designed and located, with due consideration of other implications for safety, to withstand the effects of hazards or to be protected, in accordance with their importance to safety, against hazards and against common cause failure mechanisms generated by hazards.

See also:

SSG-64 on Protection against Internal Hazards in the Design of Nuclear Power Plants

SSG-89 on Evaluation of Seismic Safety for Nuclear Installations

SSG-9 (Rev. 1) on Seismic Hazards in Site Evaluation for Nuclear Installations

SSG-18 on Meteorological and Hydrological Hazards in Site Evaluation for Nuclear Installations

SSG-21 on Volcanic Hazards in Site Evaluation for Nuclear Installations

SSG-93 on Geotechnical Aspects in the Siting and Design of Nuclear Installations

SSG-68 onDesign of Nuclear Installations Against External Events Excluding Earthquakes

SSG-67 Seismic Design for Nuclear Installations

SSG-79 Hazards Associated with Human Induced External Events in Site Evaluation for Nuclear Installations

SSG-77 , Protection Against Internal and External Hazards in the Operation of Nuclear Power Plants

5.15B. For multiple unit plant sites, the design shall take due account of the potential for specific hazards to give rise to impacts on several or even all units on the site simultaneously.

You may also wish to read:

IAEA Safety Report Series No. 96, Technical Approach to Probabilistic Safety Assessment for Multiple Reactor Units
Internal hazards
5.16. The design shall take due account of internal hazards such as fire, explosion, flooding, missile generation, collapse of structures and falling objects, pipe whip, jet impact and release of fluid from failed systems or from other installations on the site. Appropriate features for prevention and mitigation shall be provided to ensure that safety is not compromised.
External hazards
5.17 The design shall include due consideration of those natural and human induced external events¹¹11 (i.e. events of origin external to the plant) that have been identified in the site evaluation process. Causation and likelihood shall be considered in postulating potential hazards. In the short term, the safety of the plant shall not be permitted to be dependent on the availability of off-site services such as electricity supply and firefighting services. The design shall take due account of site specific conditions to determine the maximum delay time by which off-site services need to be available.

5.18. This paragraph was deleted and itscontent, with a broader scope,has been transfered to the new paragraph 5.15A.

5.19. Features shall be provided to minimize any interactions between buildings containing items important to safety (including power cabling and control cabling) and any other plant structure as a result of external events considered in the design.

5.20. This paragraph was deleted and its content, with a broader scope, has been transfered to the new paragraph 5.15A.

5.21. The design of the plant shall provide for an adequate margin to protect items important to safety against levels of external hazards to be considered for design, derived from the hazard evaluation for the site, and to avoid cliff edge effects.¹²12

5.21A. The design of the plant shall also provide for an adequate margin to protect items ultimately necessary to prevent an early radioactive release or a large radioactive release in the event of levels of natural hazards exceeding those considered for design, derived from the hazard evaluation for the site.

5.22. This paragraph was deketed and its content, with a broader scope, has been transfered to the new paragraph 5.15B.
Requirement 18: Engineering design rules
The engineering design rules for items important to safety at a nuclear power plant shall be specified and shall comply with the relevant national or international codes and standards and with proven engineering practices, with due account taken of their relevance to nuclear power technology.

5.23. Methods to ensure a robust design shall be applied, and proven engineering practices shall be adhered to in the design of a nuclear power plant to ensure that the fundamental safety functions are achieved for all operational states and for all accident conditions.
Requirement 19: Design basis accidents
A set of accidents that are to be considered in the design shall be derived from postulated initiating events for the purpose of establishing the boundary conditions for the nuclear power plant to withstand, without acceptable limits for radiation protection being exceeded.

5.24. Design basis accidents shall be used to define the design bases, including performance criteria, for safety systems and for other items important to safety that are necessary to controldesign basis accident conditions, with the objective of returning the plant to a safe state and mitigating the consequences of any accidents.

5.25. The design shall be such that for design basis accident conditions, key plant parameters do not exceed the specified designlimits. A primary objective shall be to manage all design basis accidents so that they have no, or only minor, radiological consequences, on or off the site, and do not necessitate any off-siteprotective actions.

5.26. The design basis accidents shall be analysed in a conservative manner. This approach involves postulating certain failures in safety systems, specifying design criteria and using conservative assumptions, models and input parameters in the analysis.
Requirement 20: Design extension conditions
A set of design extension conditions shall be derived on the basis of engineering judgement, deterministic assessments and probabilistic assessments for the purpose of further improving the safety of the nuclear power plant by enhancing the plant’s capabilities to withstand, without unacceptable radiological consequences, accidents that are either more severe than design basis accidents or that involve additional failures. These design extension conditions shall be used to identify the additional accidentscenarios to be addressed in the design and to plan practicable provisions for the prevention of such accidents or mitigation of their consequences.

5.27. An analysis of design extension conditions for the plant shall be performed.¹³13The main technical objective of considering the design extension conditions is to provide assurance that the design of the plant is such as to prevent accident conditions that are not considered design basis accident conditions, or to mitigate their consequences, as far as is reasonably practicable. This might require additional safety features for design extension conditions, or extension of the capability of safety systems to prevent, or to mitigate the consequences of, a severe accident, or to maintain the integrity of the containment. These additional safety features for design extension conditions, or this extension of the capability of safety systems, shall be such as to ensure the capability for managing accident conditions in which there is a significant amount of radioactive material in the containment (including radioactive material resulting from severe degradation of the reactor core). The plant shall be designed so that it can be brought into a controlled state and the containment function can be maintained, with the result that the possibility of plant states arising that could lead to an early radioactiverelease or a large radioactiverelease is ‘practically eliminated’.¹⁴14 The effectiveness of provisions to ensure the functionality of the containment could be analysed on the basis of the best estimate approach.

5.28. The design extension conditions shall be used to define the design specifications for safety features and for the design of all other items important to safety that are necessary for preventing such conditions from arising, or, if they do arise, for controlling them and mitigating their consequences.

5.29. The analysis undertaken shall include identification of the features that are designed for use in, or that are capable¹⁵15 of preventing or mitigating, events considered in the design extension conditions. These features:

Shall be independent, to the extent practicable, of those used in more frequent accidents;

Shall be capable of performing in the environmental conditions pertaining to these design extension conditions, including design extension conditions in severe accidents, where appropriate;

Shall have reliability commensurate with the function that they are required to fulfil.

5.30. In particular, the containment and its safety features shall be able to withstand extreme scenarios that include, among other things, melting of the reactor core. These scenarios shall be selected by using engineering judgement and input from probabilistic safety assessments.

5.31. The design shall be such that the possibility of conditions arising that could lead to an early radioactive release or a large radioactive release is ‘practically eliminated’.¹⁶16

5.31A. The design shall be such that for design extension conditions, protective actions that are limited in terms of lengths of time and areas of application shall be sufficient for the protection of the public, and sufficient time shall be available to take such measures.
Combinations of events and failures
5.32. Where the results of engineering judgement, deterministic safety assessments and probabilistic safety assessments indicate that combinations of events could lead to anticipated operational occurrences or to accident conditions, such combinations of events shall be considered to be design basis accidents or shall be included as part of design extension conditions, depending mainly on their likelihood of occurrence. Certain events might be consequences of other events, such as a flood following an earthquake. Such consequential effects shall be considered to be part of the original postulated initiating event.
Requirement 21: Physical separation and independence of safety systems
Interference between safety systems or between redundant elements of a system shall be prevented by means such as physical separation, electrical isolation, functional independence and independence of communication (data transfer), as appropriate.

5.33. Safety system equipment (including cables and raceways) shall be readily identifiable in the plant for each redundant element of a safety system.
Requirement 22: Safety classification
All items important to safety shall be identified and shall be classified on the basis of their function and their safety significance.

5.34. The method for classifying the safety significance of items important to safety shall be based primarily on deterministic methods complemented, where appropriate, by probabilistic methods, with due account taken of factors such as:

The safety function(s) to be performed by the item;

The consequences of failure to perform a safety function;

The frequency with which the item will be called upon to perform a safety function;

The time following a postulated initiating event at which, or the period for which, the item will be called upon to perform a safety function.

5.35. The design shall be such as to ensure that any interference between items important to safety will be prevented, and in particular that any failure of items important to safety in a system in a lower safety class will not propagate to a system in a higher safety class.

See also SSG-30 on Safety Classification of Structures, Systems and Components in Nuclear Power Plants

You may also read the TECDOC No. 1787 on Application of the Safety Classification of Structures, Systems and Components in Nuclear Power Plants

5.36. Equipment that performs multiple functions shall be classified in a safety class that is consistent with the most important function performed by the equipment.
Requirement 23: Reliability of items important to safety
The reliability of items important to safety shall be commensurate with their safety significance.

5.37. The design of items important to safety shall be such as to ensure that the equipment can be qualified, procured, installed, commissioned, operated and maintained to be capable of withstanding, with sufficient reliability and effectiveness, all conditions specified in the design basis for the items.

5.38. In the selection of equipment, consideration shall be given to both spurious operation and unsafe failure modes. Preference shall be given in the selection process to equipment that exhibits a predictable and revealed mode of failure and for which the design facilitates repair or replacement.
Requirement 24: Common cause failures
The design of equipment shall take due account of the potential for common cause failures of items important to safety, to determine how the concepts of diversity, redundancy, physical separation and functional independence have to be applied to achieve the necessary reliability.

.
Requirement 25: Single failure criterion
The single failure criterion shall be applied to each safety group incorporated in the plant design.

5.39 Spurious action shall be considered to be one mode of failure when applying the single failure criterion¹⁷17 to a safety group or safety system.

5.40. The design shall take due account of the failure of a passive component, unless it has been justified in the single failureanalysis with a high level of confidence that a failure of that component is very unlikely and that its function would remain unaffected by the postulated initiating event.
Requirement 26: Fail-safe design
The concept of fail-safe design shall be incorporated, as appropriate, into the design of systems and components important to safety.

5.41. Systems and components important to safety shall be designed for fail-safe behaviour, as appropriate, so that their failure or the failure of a support feature does not prevent the performance of the intended safety function.
Requirement 27: Support service systems
Support service systems that ensure the operability of equipment forming part of a system important to safety shall be classified accordingly.

5.42. The reliability, redundancy, diversity and independence of support service systems and the provision of features for their isolation and for testing their functional capability shall be commensurate with the significance to safety of the system being supported.

5.43. It shall not be permissible for a failure of a support service system to be capable of simultaneously affecting redundant parts of a safety system or a system fulfilling diverse safety functions and compromising the capability of these systems to fulfil their safety functions.
Requirement 28: Operational limits and conditions for safe operation
The design shall establish a set of operational limits and conditions for safe operation of the nuclear power plant.

5.44. The requirements and operational limits and conditions established in the design for the nuclear power plant shall include (Requirement 6 of IAEA Safety Standards Series No. SSR-2/2 (Rev. 1), Safety of Nuclear Power Plants: Commissioning and Operation [4]):

Safety limits;

Limiting settings for safety systems;

Limits and conditions for normal operation;

Control system constraints and procedural constraints on process variables and other important parameters;

Requirements for surveillance, maintenance, testing and inspection of the plant to ensure that structures, systems and components function as intended in the design, to comply with the requirement for optimization by keeping radiation risksas low as reasonably achievable;

Specified operational configurations, including operational restrictions in the event of the unavailability of safety systems or safety related systems;

Action statements, including completion times for actions in response to deviations from the operational limits and conditions.

See also SSR-2/2 (Rev. 1) on Safety of Nuclear Power Plants: Commissioning and Operation Requirements 6 and SSG-70 on Operational Limits and Conditions and Operating Procedures for Nuclear Power Plants
Design for safe operation over the lifetime of the plant Requirement 29: Calibration, testing, maintenance, repair, replacement, inspection and monitoring of items important to safety
Items important to safety for a nuclear power plant shall be designed to be calibrated, tested, maintained, repaired or replaced, inspected and monitored as required to ensure their capability of performing their functions and to maintain their integrity in all conditions specified in their design basis.

5.45. The plant layout shall be such that activities for calibration, testing, maintenance, repair or replacement, inspection and monitoring are facilitated and can be performed to relevant national and international codes and standards. Such activities shall be commensurate with the importance of the safety functions to be performed, and shall be performed without undue exposure of workers.

See also SSR-2/2 (Rev. 1) on Safety of Nuclear Power Plants: Commissioning and Operation Requirements 9 and 31, SSG-74 on Maintenance, Surveillance and In-Service Inspection in Nuclear Power Plants and SSG-76 on Conduct of Operations at Nuclear Power Plants

5.46. Where items important to safety are planned to be calibrated, tested or maintained during power operation, the respective systems shall be designed for performing such tasks with no significant reduction in the reliability of performance of the safety functions. Provisions for calibration, testing, maintenance, repair, replacement or inspection of items important to safety during shutdown shall be included in the design so that such tasks can be performed with no significant reduction in the reliability of performance of the safety functions.

5.47. If an item important to safety cannot be designed to be capable of being tested, inspected or monitored to the extent desirable, a robust technical justification shall be provided that incorporates the following approach:

Other proven alternative and/or indirect methods such as surveillance testing of reference items or use of verified and validated calculational methods shall be specified.

Conservative safety margins shall be applied or other appropriate precautions shall be taken to compensate for possible unanticipated failures.

Requirement 30: Qualification of items important to safety
A qualification programme for items important to safety shall be implemented to verify that items important to safety at a nuclear power plant are capable of performing their intended functions when necessary, and in the prevailing environmental conditions, throughout their design life, with due account taken of plant conditions during maintenance and testing.

5.48. The environmental conditions considered in the qualification programme for items important to safety at a nuclear power plant shall include the variations in ambient environmental conditions that are anticipated in the design basis for the plant.

5.49. The qualification programme for items important to safety shall include the consideration of ageing effects caused by environmental factors (such as conditions of vibration, irradiation, humidity or temperature) over the expected service life of the items important to safety. When the items important to safety are subject to natural external events and are required to perform a safety function during or following such an event, the qualification programme shall replicate as far as is practicable the conditions imposed on the items important to safety by the natural external event, either by test or analysis, or by a combination of both.

See also

SSG-69 Equipment Qualification for Nuclear Installations

SSG-30 on Safety Classification of Structures, Systems and Components in Nuclear Power Plants

SSG-67 Seismic Design for Nuclear Installations

You may also wish to read:

IAEA-TECDOC-1818 , Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions

5.50. Any environmental conditions that could reasonably be anticipated and that could arise in specific operational states, such as in periodic testing of the containment leak rate, shall be included in the qualification programme.
Requirement 31: Ageing management
The design life of items important to safety at a nuclear power plant shall be determined. Appropriate margins shall be provided in the design to take due account of relevant mechanisms of ageing, neutron embrittlement and wear out and of the potential for age related degradation, to ensure the capability of items important to safety to perform their necessary safety functions throughout their design life.

5.51. The design for a nuclear power plant shall take due account of ageing and wear out effects in all operational states for which a component is credited, including testing, maintenance, maintenance outages, plant states during a postulated initiating event and plant states following a postulated initiating event.

5.52. Provision shall be made for monitoring, testing, sampling and inspection to assess ageing mechanisms predicted at the design stage and to help to identify unanticipated behaviour of the plant or degradation that might occur in service.

See also SSR-2/2 (Rev. 1) on Safety of Nuclear Power Plants: Commissioning and Operation Requirements 14 and SSG-48 on Ageing Management and Development of a Programme for Long Term Operation of Nuclear Power Plants
Human factors Requirement 32: Design for optimal operator performance
Systematic consideration of human factors, including the human–machine interface, shall be included at an early stage in the design process for a nuclear power plant and shall be continued throughout the entire design process.

5.53. The design for a nuclear power plant shall specify the minimum number of operating personnel required to perform all the simultaneous operations necessary to bring the plant into a safe state.

5.54. Operating personnel who have gained operating experience in similar plants shall, as far as is practicable, be actively involved in the design process conducted by the design organization, in order to ensure that consideration is given as early as possible in the process to the future operation and maintenance of equipment.

5.55. The design shall support operating personnel in the fulfilment of their responsibilities and in the performance of their tasks, and shall limit the likelihood and the effects of operating errors on safety. The design process shall give due consideration to plant layout and equipment layout, and to procedures, including procedures for maintenance and inspection, to facilitate interaction between the operating personnel and the plant, in all plant states.

5.56. The human–machine interface shall be designed to provide the operators with comprehensive but easily manageable information, in accordance with the necessary decision times and action times. The information necessary for the operator to make decisions to act shall be simply and unambiguously presented.

5.57. The operator shall be provided with the necessary information:

To assess the general state of the plant in any condition;

To operate the plant within the specified limits on parameters associated with plant systems and equipment (operational limits and conditions);

To confirm that safety actions for the actuation of safety systems are automatically initiated when needed and that the relevant systems perform as intended;

To determine both the need for and the time for manual initiation of the specified safety actions.

5.58. The design shall be such as to promote the success of operator actions with due regard for the time available for action, the conditions to be expected and the psychological demands being made on the operator.

5.59. The need for intervention by the operator on a short time scale shall be kept to a minimum, and it shall be demonstrated that the operator has sufficient time to make a decision and sufficient time to act.

5.60. The design shall be such as to ensure that, following an event affecting the plant, environmental conditions in the control room or the supplementary control room and in locations on the access route to the supplementary control room do not compromise the protection and safety of the operating personnel.

5.61. The design of workplaces and the working environment of the operating personnel shall be in accordance with ergonomic concepts.

5.62. Verification and validation, including by the use of simulators, of features relating to human factors shall be included at appropriate stages to confirm that necessary actions by the operator have been identified and can be correctly performed.

See also:

SSG-51 Human Factors Engineering in the Design of Nuclear Power Plants
Other design considerations Requirement 33: Safety systems, and safety features for design extension conditions, of units of a multiple unit nuclear power plant
Each unit of a multiple unit nuclear power plant shall have its own safety systems and shall have its own safety features for design extension conditions.

5.63. To further enhance safety, means allowing interconnections between units of a multiple unit nuclear power plant shall be considered in the design.

You may also wish to read:

IAEA Safety Report Series No. 96, Technical Approach to Probabilistic Safety Assessment for Multiple Reactor Units
Requirement 34: Systems containing fissile material or radioactive material
All systems in a nuclear power plant that could contain fissile material or radioactive material shall be so designed as: to prevent the occurrence of events that could lead to an uncontrolled radioactiverelease to the environment; to prevent accidental criticality and overheating; to ensure that radioactivereleases are kept below authorized limits on discharges in normal operation and below acceptable limits in accident conditions, and are kept as low as reasonably achievable; and to facilitate mitigation of radiological consequences of accidents.

See also SSG-27 (Rev. 1) on Criticality Safety in the Handling of Fissile Material
Requirement 35: Nuclear power plants used for cogeneration of heat and power, heat generation or desalination
Nuclear power plants coupled with heat utilization units (such as for district heating) and/or water desalination units shall be designed to prevent processes that transport radionuclides from the nuclear plant to the desalination unit or the district heating unit under conditions of operational states and in accident conditions.

.
Requirement 36: Escape routes from the plant
A nuclear power plant shall be provided with a sufficient number of escape routes, clearly and durably marked, with reliable emergency lighting, ventilation and other services essential to the safe use of these escape routes.

5.64. Escape routes from the nuclear power plant shall meet the relevant national and international requirements for radiation zoning and fire protection, and the relevant national requirements for industrial safety and plant security.

5.65. At least one escape route shall be available from workplaces and other occupied areas following an internal event or an external event or following combinations of events considered in the design.
Requirement 37: Communication systems at the plant
Effective means of communication shall be provided throughout the nuclear power plant to facilitate safe operation in all modes of normal operation and to be available for use following all postulated initiating events and in accident conditions.

5.66. Suitable alarm systems and means of communication shall be provided so that all persons present at the nuclear power plant and on the site can be given warnings and instructions, in operational states and in accident conditions.

5.67. Suitable and diverse means of communication necessary for safety within the nuclear power plant and in the immediate vicinity, and for communication with relevant off-site agencies, shall be provided.
Requirement 38: Control of access to the plant
The nuclear power plant shall be isolated from its surroundings with a suitable layout of the various structural elements so that access to it can be controlled.

5.68. Provision shall be made in the design of the buildings and the layout of the site for the control of access to the nuclear power plant by operating personnel and/or for equipment, including emergency response personnel and vehicles, with particular consideration given to guarding against the unauthorized entry of persons and goods to the plant.

See also:

NSS No. 13 Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities (INFCIRC/225/Revision 5)

NSS No. 8-G (Rev. 1) Preventive and Protective Measures against Insider Threats
Requirement 39: Prevention of unauthorized access to, or interference with, items important to safety
Unauthorized access to, or interference with, items important to safety, including computer hardware and software, shall be prevented.

See also:

NSS No. 13 Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities (INFCIRC/225/Revision 5)

NSS No. 8-G (Rev. 1) Preventive and Protective Measures against Insider Threats
Requirement 40: Prevention of harmful interactions of systems important to safety
The potential for harmful interactions of systems important to safety at the nuclear power plant that might be required to operate simultaneously shall be evaluated, and effects of any harmful interactions shall be prevented.

5.69. In the analysis of the potential for harmful interactions of systems important to safety, due account shall be taken of physical interconnections and of the possible effects of one system’s operation, maloperation or malfunction on local environmental conditions of other essential systems, to ensure that changes in environmental conditions do not affect the reliability of systems or components in functioning as intended.

5.70. If two fluid systems important to safety are interconnected and are operating at different pressures, either the systems shall both be designed to withstand the higher pressure, or provision shall be made to prevent the design pressure of the system operating at the lower pressure from being exceeded.
Requirement 41: Interactions between the electrical power grid and the plant
The functionality of items important to safety at the nuclear power plant shall not be compromised by disturbances in the electrical power grid, including anticipated variations in the voltage and frequency of the grid supply.

.
Safety analysis Requirement 42: Safety analysis of the plant design
A safety analysis of the design for the nuclear power plant shall be conducted in which methods of both deterministic analysis and probabilistic analysis shall be applied to enable the challenges to safety in the various categories of plant states to be evaluated and assessed.

5.71. On the basis of a safety analysis, the design basis for items important to safety and their links to initiating events and event sequences shall be confirmed.¹⁸18It shall be demonstrated that the nuclear power plant as designed is capable of complying with authorized limits on discharges with regard to radioactivereleases and with the dose limits in all operational states, and is capable of meeting acceptable limits for accident conditions.

5.72. The safety analysis shall provide assurance that defence in depth has been implemented in the design of the plant.
5.73.
The safety analysis shall provide assurance that uncertainties have been given adequate consideration in the design of the plant and in particular that adequate margins are available to avoid cliff edge effects¹⁹19 and early radioactive releases or large radioactive releases.

5.74. The applicability of the analytical assumptions, methods and degree of conservatism used in the design of the plant shall be updated and verified for the current or as built design.

See also GSR Part 4 (Rev. 1) on Safety Assessment for Facilities and Activities, SSG-2 (Rev. 1) on Deterministic Safety Analysis for Nuclear Power Plants, SSG-3 (Rev. 1) on Development and Application of Level 1 Probabilistic Safety Assessment for Nuclear Power Plants and SSG-4 (Rev. 1) on Development and Application of Level 2 Probabilistic Safety Assessment for Nuclear Power Plants
Deterministic approach
5.75. The deterministic safety analysis shall mainly provide:

Establishment and confirmation of the design bases for all items important to safety;

Characterization of the postulated initiating events that are appropriate for the site and the design of the plant;

Analysis and evaluation of event sequences that result from postulated initiating events, to confirm the qualificationrequirements;

Comparison of the results of the analysis with acceptance criteria, designlimits, dose limits and acceptable limits for purposes of radiation protection;

Demonstration that the management of anticipated operational occurrences and design basis accident conditions is possible by safety actions for the automatic actuation of safety systems in combination with prescribed actions by the operator;

Demonstration that the management of design extension conditions is possible by the automatic actuation of safety systems and the use of safety features in combination with expected actions by the operator.

Probabilistic approach
5.76. The design shall take due account of the probabilistic safety analysis of the plant for all modes of operation and for all plant states, including shutdown, with particular reference to:

Establishing that a balanced design has been achieved such that no particular feature or postulated initiating event makes a disproportionately large or significantly uncertain contribution to the overall risks, and that, to the extent practicable, the levels of defence in depth are independent;

Providing assurance that situations in which small deviations in plant parameters could give rise to large variations in plant conditions (cliff edge effects) will be prevented;²⁰20

Comparing the results of the analysis with the acceptance criteria for risk where these have been specified.

6. Design of specific plant systems Reactor core and associated features Requirement 43: Performance of fuel elements and assemblies
Fuel elements and assemblies for the nuclear power plant shall be designed to maintain their structural integrity, and to withstand satisfactorily the anticipated radiation levels and other conditions in the reactor core, in combination with all the processes of deterioration that could occur in operational states.

6.1. The processes of deterioration to be considered shall include those arising from:

Differential expansion and deformation;

External pressure of the coolant;

Additional internal pressure due to fission products and the buildup of helium in fuel elements;

Irradiation of fuel and other materials in the fuel assembly;

Variations in pressure and temperature resulting from variations in power demand;

Chemical effects;

Static and dynamic loading, including flow induced vibrations and mechanical vibrations;

Variations in performance in relation to heat transfer that could result from distortion or chemical effects.

Allowance shall be made for uncertainties in data, in calculations and in manufacture.

6.2. Fuel design limits shall include limits on the permissible leakage of fission products from the fuel in anticipated operational occurrences so that the fuel remains suitable for continued use.

6.3. Fuel elements and fuel assemblies shall be capable of withstanding the loads and stresses associated with fuel handling.

See also SSG-52 on Design of the Reactor Core for Nuclear Power Plants
Requirement 44: Structural capability of the reactor core
The fuel elements and fuel assemblies and their supporting structures for the nuclear power plant shall be designed so that, in operational states and in accident conditions other than severe accidents, a geometry that allows for adequate cooling is maintained and the insertion of control rods is not impeded.

See also SSG-52 on Design of the Reactor Core for Nuclear Power Plants
Requirement 45: Control of the reactor core
Distributions of neutron flux that can arise in any state of the reactor core in the nuclear power plant, including states arising after shutdown and during or after refuelling, and states arising from anticipated operational occurrences and from accident conditions not involving degradation of the reactor core, shall be inherently stable. The demands made on the control system for maintaining the shapes, levels and stability of the neutron flux within specified designlimits in all operational states shall be minimized.

6.4. Adequate means of detecting the neutron flux distributions in the reactor core and their changes shall be provided for the purpose of ensuring that there are no regions of the core in which the designlimits could be exceeded.

6.5. In the design of reactivitycontrol devices, due account shall be taken of wear out and of the effects of irradiation, such as burnup, changes in physical properties and production of gas.

6.6. The maximum degree of positive reactivity and its rate of increase by insertion in operational states and accident conditions not involving degradation of the reactor core shall be limited or compensated for, to prevent any resultant failure of the pressure boundary of the reactor coolant systems, to maintain the capability for cooling and to prevent any significant damage to the reactor core.

See also SSG-52 on Design of the Reactor Core for Nuclear Power Plants
Requirement 46: Reactor shutdown
Means shall be provided to ensure that there is a capability to shut down the reactor of the nuclear power plant in operational states and in accident conditions, and that the shutdowncondition can be maintained even for the most reactive conditions of the reactor core.

6.7. The effectiveness, speed of action and shutdown margin of the means of shutdown of the reactor shall be such that the specified designlimits for fuel are not exceeded.

6.8. In judging the adequacy of the means of shutdown of the reactor, consideration shall be given to failures arising anywhere in the plant that could render part of the means of shutdown inoperative (such as failure of a control rod to insert) or that could result in a common cause failure.

6.9. The means for shutting down the reactor shall consist of at least two diverse and independent systems.

6.10. At least one of the two different shutdown systems shall be capable, on its own, of maintaining the reactor subcritical by an adequate margin and with high reliability, even for the most reactive conditions of the reactor core.

6.11. The means of shutdown shall be adequate to prevent any foreseeable increase in reactivity leading to unintentional criticality during the shutdown, or during refuelling operations or other routine or non-routine operations in the shutdown state.

6.12. Instrumentation shall be provided and tests shall be specified for ensuring that the means of shutdown are always in the state stipulated for a given plant state.

See also SSG-52 on Design of the Reactor Core for Nuclear Power Plants
Reactor coolant systems Requirement 47: Design of reactor coolant systems
The components of the reactor coolant systems for the nuclear power plant shall be designed and constructed so that the risk of faults due to inadequate quality of materials, inadequate design standards, insufficient capability for inspection or inadequate quality of manufacture is minimized.

6.13. Pipework connected to the pressure boundary of the reactor coolant systems for the nuclear power plant shall be equipped with adequate isolation devices to limit any loss of radioactive fluid (primary coolant) and to prevent the loss of coolant through interfacing systems.

6.14. The design of the reactor coolant pressure boundary shall be such that flaws are very unlikely to be initiated, and any flaws that are initiated would propagate in a regime of high resistance to unstable fracture and to rapid crack propagation, thereby permitting the timely detection of flaws.

6.15. The design of the reactor coolant systems shall be such as to ensure that plant states in which components of the reactor coolant pressure boundary could exhibit embrittlement are avoided.

6.16. The design of the components contained inside the reactor coolant pressure boundary, such as pump impellers and valve parts, shall be such as to minimize the likelihood of failure and consequential damage to other components of the primary coolant system that are important to safety, in all operational states and in design basis accident conditions, with due allowance made for deterioration that might occur in service.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 48: Overpressure protection of the reactor coolant pressure boundary
Provision shall be made to ensure that the operation of pressure relief devices will protect the pressure boundary of the reactor coolant systems against overpressure and will not lead to the release of radioactive material from the nuclear power plant directly to the environment.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 49: Inventory of reactor coolant
Provision shall be made for controlling the inventory, temperature and pressure of the reactor coolant to ensure that specified designlimits are not exceeded in any operational state of the nuclear power plant, with due account taken of volumetric changes and leakage.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 50: Cleanup of reactor coolant
Adequate facilities shall be provided at the nuclear power plant for the removal from the reactor coolant of radioactive substances, including activated corrosion products and fission products deriving from the fuel, and non-radioactive substances.

6.17. The capabilities of the necessary plant systems shall be based on the specified designlimit on permissible leakage of the fuel, with a conservative margin to ensure that the plant can be operated with a level of circuit activity that is as low as reasonably practicable, and to ensure that the requirements are met for radioactivereleases to be as low as reasonably achievable and below the authorized limits on discharges.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 51: Removal of residual heat from the reactor core
Means shall be provided for the removal of residual heat from the reactor core in the shutdown state of the nuclear power plant such that the designlimits for fuel, the reactor coolant pressure boundary and structures important to safety are not exceeded.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 52: Emergency cooling of the reactor core
Means of cooling the reactor core shall be provided to restore and maintain cooling of the fuel under accident conditions at the nuclear power plant, even if the integrity of the pressure boundary of the primary coolant system is not maintained.

6.18. The means provided for cooling of the reactor core shall be such as to ensure that:

The limiting parameters for the cladding or for integrity of the fuel (such as temperature) will not be exceeded;

Possible chemical reactions are kept to an acceptable level;

The effectiveness of the means of cooling of the reactor core compensates for possible changes in the fuel and in the internal geometry of the reactor core;

Cooling of the reactor core will be ensured for a sufficient time.

6.19. Design features (such as leak detection systems, appropriate interconnections and capabilities for isolation) and suitable redundancy and diversity shall be provided to fulfil the requirements of para. 6.18 with adequate reliability for each postulated initiating event.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 53: Heat transfer to an ultimate heat sink
The capability to transfer heat to an ultimate heat sink shall be ensured for all plant states.

6.19A. Systems for transferring heat shall have adequate reliability for the plant states in which they have to fulfil the heat transfer function. This may require the use of a different ultimate heat sink or different access to the ultimate heat sink.

6.19B. The heat transfer function shall be fulfilled for levels of natural hazards more severe than those considered for design, derived from the hazard evaluation for the site.

See also SSG-56 : Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Containment structure and containment system Requirement 54: Containment system for the reactor
A containment system shall be provided to ensure, or to contribute to, the fulfilment of the following safety functions at the nuclear power plant: (i) confinement of radioactive substances in operational states and in accident conditions; (ii) protection of the reactor against natural external events and human induced events; and (iii) radiation shielding in operational states and in accident conditions.

See also SSG-53Design of the Reactor Containment and Associated Systems for Nuclear Power Plants
Requirement 55: Control of radioactive releases from the containment
The design of the containment shall be such as to ensure that any radioactiverelease from the nuclear power plant to the environment is as low as reasonably achievable, is below the authorized limits on discharges in operational states and is below acceptable limits in accident conditions.

6.20. The containmentstructure and the systems and components affecting the leaktightness of the containment system shall be designed and constructed so that the leak rate can be tested after all penetrations through the containment have been installed and, if necessary, during the operating lifetime of the plant, so that the leak rate can be tested at the containmentdesign pressure.

6.21. The number of penetrations through the containment shall be kept to a practical minimum and all penetrations shall meet the same designrequirements as the containmentstructure itself. The penetrations shall be protected against reaction forces caused by pipe movement or accidental loads such as those due to missiles caused by external or internal events, jet forces and pipe whip.

See also SSG-53Design of the Reactor Containment and Associated Systems for Nuclear Power Plants
Requirement 56: Isolation of the containment
Each line that penetrates the containment at a nuclear power plant as part of the reactor coolant pressure boundary or that is connected directly to the containment atmosphere shall be automatically and reliably sealable in the event of an accident in which the leaktightness of the containment is essential to preventing radioactivereleases to the environment that exceed acceptable limits.

6.22. Lines that penetrate the containment as part of the reactor coolant pressure boundary and lines that are connected directly to the containment atmosphere shall be fitted with at least two adequate containmentisolation valves or check valves arranged in series²¹21 and shall be provided with suitable leak detection systems. Containmentisolation valves or check valves shall be located as close to the containment as is practicable, and each valve shall be capable of reliable and independent actuation and of being periodically tested.

6.23. Exceptions to the requirements for containmentisolation stated in para. 6.22 shall be permissible for specific classes of lines such as instrumentation lines, or in cases in which application of the methods of containmentisolation specified in para. 6.22 would reduce the reliability of a safety system that includes a penetration of the containment.

6.24. Each line that penetrates the containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one adequate containmentisolation valve. The containmentisolation valves shall be located outside the containment and as close to the containment as is practicable.

See also SSG-53Design of the Reactor Containment and Associated Systems for Nuclear Power Plants
Requirement 57: Access to the containment
Access by operating personnel to the containment at a nuclear power plant shall be through airlocks equipped with doors that are interlocked to ensure that at least one of the doors is closed during reactor power operation and in accident conditions.

6.25. Where provision is made for entry of operating personnel for surveillance purposes, provision for ensuring protection and safety for operating personnel shall be specified in the design. Where equipment airlocks are provided, provision for ensuring protection and safety for operating personnel shall be specified in the design.

6.26. Containment openings for the movement of equipment or material through the containment shall be designed to be closed quickly and reliably in the event that isolation of the containment is required.

See also SSG-53 Design of the Reactor Containment and Associated Systems for Nuclear Power Plants
Requirement 58: Control of containment conditions
Provision shall be made to control the pressure and temperature in the containment at a nuclear power plant and to control any buildup of fission products or other gaseous, liquid or solid substances that might be released inside the containment and that could affect the operation of systems important to safety.

6.27. The design shall provide for sufficient flow routes between separate compartments inside the containment. The cross-sections of openings between compartments shall be of such dimensions as to ensure that the pressure differentials occurring during pressure equalization in accident conditions do not result in unacceptable damage to the pressure bearing structure or to systems that are important in mitigating the effects of accident conditions.

6.28. The capability to remove heat from the containment shall be ensured, in order to reduce the pressure and temperature in the containment, and to maintain them at acceptably low levels after any accidental release of high energy fluids. The systems performing the function of removal of heat from the containment shall have sufficient reliability and redundancy to ensure that this function can be fulfilled.

6.28A. Design provision shall be made to prevent the loss of the structural integrity of the containment in all plant states. The use of this provision shall not lead to an early radioactive release or a large radioactive release.

6.28B. The design shall also include features to enable the safe use of non-permanent equipment²²22 for restoring the capability to remove heat from the containment.

6.29. Design features to controlfission products, hydrogen, oxygen and other substances that might be released into the containment shall be provided as necessary:

To reduce the amounts of fission products that could be released to the environment in accident conditions;

To control the concentrations of hydrogen, oxygen and other substances in the containment atmosphere in accident conditions so as to prevent deflagration or detonation loads that could challenge the integrity of the containment.

6.30. Coverings, thermal insulations and coatings for components and structures within the containment system shall be carefully selected and methods for their application shall be specified to ensure the fulfilment of their safety functions and to minimize interference with other safety functions in the event of deterioration of the coverings, thermal insulations and coatings.

See also SSG-53 Design of the Reactor Containment and Associated Systems for Nuclear Power Plants
Instrumentation and control systems Requirement 59: Provision of instrumentation
Instrumentation shall be provided for: determining the values of all the main variables that can affect the fission process, the integrity of the reactor core, the reactor coolant systems and the containment at the nuclear power plant; for obtaining essential information on the plant that is necessary for its safe and reliable operation; for determining the status of the plant in accident conditions; and for making decisions for the purposes of accident management.

6.31. Instrumentation and recording equipment shall be provided to ensure that essential information is available for monitoring the status of essential equipment and the course of accidents, for predicting the locations of releases and the amounts of radioactive material that could be released from the locations that are so intended in the design, and for post-accident analysis.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 60: Control systems
Appropriate and reliable control systems shall be provided at the nuclear power plant to maintain and limit the relevant process variables within the specified operational ranges.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 61: Protection system
A protection system shall be provided at the nuclear power plant that has the capability to detect unsafe plant conditions and to initiate safety actions automatically to actuate the safety systems necessary for achieving and maintaining safe plant conditions.

6.32. The protection system shall be designed:

To be capable of overriding unsafe actions of the control system;

With fail-safe characteristics to achieve safe plant conditions in the event of failure of the protection system.

6.33. The design:

Shall prevent operator actions that could compromise the effectiveness of the protection system in operational states and in accident conditions, but shall not counteract correct operator actions in accident conditions;

Shall automate various safety actions to actuate safety systems so that operator action is not necessary within a justified period of time from the onset of anticipated operational occurrences or accident conditions;

Shall make relevant information available to the operator for monitoring the effects of automatic actions.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 62: Reliability and testability of instrumentation and control systems
Instrumentation and control systems for items important to safety at the nuclear power plant shall be designed for high functional reliability and periodic testability commensurate with the safety function(s) to be performed.

6.34. Design techniques such as testability, including a self-checking capability where necessary, fail-safe characteristics, functional diversity and diversity in componentdesign and in concepts of operation shall be used to the extent practicable to prevent the loss of a safety function.

6.35. Safety systems shall be designed to permit periodic testing of their functionality when the plant is in operation, including the possibility of testing channels independently for the detection of failures and losses of redundancy. The design shall permit all aspects of functionality testing for the sensor, the input signal, the final actuator and the display.

6.36. When a safety system, or part of a safety system, has to be taken out of service for testing, adequate provision shall be made for the clear indication of any protection systembypasses that are necessary for the duration of the testing or maintenanceactivities.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 63: Use of computer based equipment in systems important to safety
If a system important to safety at the nuclear power plant is dependent upon computer based equipment, appropriate standards and practices for the development and testing of computer hardware and software shall be established and implemented throughout the service life of the system, and in particular throughout the software development cycle. The entire development shall be subject to a quality management system.

6.37. For computer based equipment in safety systems or safety related systems:

A high quality of, and best practices for, hardware and software shall be used, in accordance with the importance of the system to safety;

The entire development process, including control, testing and commissioning of design changes, shall be systematically documented and shall be reviewable;

An assessment of the equipment shall be undertaken by experts who are independent of the design team and the supplier team to provide assurance of its high reliability;

Where safety functions are essential for achieving and maintaining safe conditions, and the necessary high reliability of the equipment cannot be demonstrated with a high level of confidence, diverse means of ensuring fulfilment of the safety functions shall be provided;

Common cause failures deriving from software shall be taken into consideration;

Protection shall be provided against accidental disruption of, or deliberate interference with, system operation.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 64: Separation of protection systems and control systems
Interference between protection systems and control systems at the nuclear power plant shall be prevented by means of separation, by avoiding interconnections or by suitable functional independence.

6.38. If signals are used in common by both a protection system and any control system, separation (such as by adequate decoupling) shall be ensured and the signal system shall be classified as part of the protection system.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 65: Control room
A control room shall be provided at the nuclear power plant from which the plant can be safely operated in all operational states, either automatically or manually, and from which measures can be taken to maintain the plant in a safe state or to bring it back into a safe state after anticipated operational occurrences and accident conditions.

6.39. Appropriate measures shall be taken, including the provision of barriers between the control room at the nuclear power plant and the external environment, and adequate information shall be provided for the protection of occupants of the control room, for a protracted period of time, against hazards such as high radiation levels resulting from accident conditions, releases of radioactive material, fire, or explosive or toxic gases.

6.40. Special attention shall be paid to identifying those events, both internal and external to the control room, that could challenge its continued operation, and the design shall provide for reasonably practicable measures to minimize the consequences of such events.

6.40A. The design of the control room shall provide an adequate margin against levels of natural hazards more severe than those considered for design, derived from the hazard evaluation for the site.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 66: Supplementary control room
Instrumentation and control equipment shall be kept available, preferably at a single location (a supplementary control room) that is physically, electrically and functionally separate from the control room at the nuclear power plant. The supplementary control room shall be so equipped that the reactor can be placed and maintained in a shutdown state, residual heat can be removed, and essential plant variables can be monitored if there is a loss of ability to perform these essential safety functions in the control room.

6.41. The requirements of para. 6.39 for taking appropriate measures and providing adequate information for the protection of occupants against hazards also apply for the supplementary control room at the nuclear power plant.

See also:

SSG-39 : Design of Instrumentation and Control Systems for Nuclear Power Plants
Requirement 67: Emergency response facilities on the site
The nuclear power plant shall include the necessary emergency response facilitieson the site. Their design shall be such that personnel will be able to perform expected tasks for managing an emergency under conditions generated by accidents and hazards.

6.42 Information about important plant parameters and radiological conditions at the nuclear power plant and in its immediate surroundings shall be provided to the relevant emergency response facilities²³23. Each facility shall be provided with means of communication with, as appropriate, the control room, the supplementary control room and other important locations at the plant, and with on-site and off-siteemergencyresponse organizations.

See also GSR Part 7 on Preparedness and Response for a Nuclear or Radiological Emergency, GS-G-2.1 on Arrangements for Preparedness for a Nuclear or Radiological Emergency, and SSG-39 Design of Instrumentation and Control Systems for Nuclear Power Plants
Emergency power supply Requirement 68: Design for withstanding the loss of off-site power
The design of the nuclear power plant shall include an emergency power supply capable of supplying the necessary power in anticipated operational occurrences and design basis accidents, in the event of a loss of off-site power. The design shall include an alternate power source to supply the necessary power in design extension conditions.

6.43. The design specifications for the emergency power supply and for the alternate power source at the nuclear power plant shall include the requirements for capability, availability, duration of the required power supply, capacity and continuity.

See also SSG-34 on Design of Electrical Power Systems for Nuclear Power Plants

6.44. The combined means to provide emergency power (such as water, steam or gas turbines, diesel engines or batteries) shall have a reliability and type that are consistent with all the requirements of the safety systems to be supplied with power, and their functional capability shall be testable.

6.44A. The alternate power source shall be capable of supplying the necessary power to preserve the integrity of the reactor coolant system and to prevent significant damage to the core and to spent fuel in the event of the loss of off-site power combined with failure of the emergency power supply.

6.44B. Equipment that is necessary to mitigate the consequences of melting of the reactor core shall be capable of being supplied by any of the available power sources.

6.44C. The alternate power source shall be independent of and physically separated from the emergency power supply. The connection time of the alternate power source shall be consistent with the depletion time of the battery.

6.44D. Continuity of power for the monitoring of the key plant parameters and for the completion of short term actions necessary for safety shall be maintained in the event of loss of the AC (alternating current) power sources.

6.45. The design basis for any diesel engine or other prime mover²⁴24 that provides an emergency power supply to items important to safety shall include:

The capability of the associated fuel oil storage and supply systems to satisfy the demand within the specified time period;

The capability of the prime mover to start and to function successfully under all specified conditions and at the required time;

Auxiliary systems of the prime mover, such as coolant systems.

6.45A. The design shall also include features to enable the safe use of non-permanent equipment to restore the necessary electrical power supply.²⁵25
Supporting systems and auxiliary systems Requirement 69: Performance of supporting systems and auxiliary systems
The design of supporting systems and auxiliary systems shall be such as to ensure that the performance of these systems is consistent with the safety significance of the system or component that they serve at the nuclear power plant.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants
Requirement 70: Heat transport systems
Auxiliary systems shall be provided as appropriate to remove heat from systems and components at the nuclear power plant that are required to function in operational states and in accident conditions.

6.46. The design of heat transport systems shall be such as to ensure that non-essential parts of the systems can be isolated.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants

SSG-56 Design of the Reactor Coolant System and Associated Systems for Nuclear Power Plants
Requirement 71: Process sampling systems and post-accident sampling systems
Process sampling systems and post-accident sampling systems shall be provided for determining, in a timely manner, the concentration of specified radionuclides in fluid process systems, and in gas and liquid samples taken from systems or from the environment, in all operational states and in accident conditions at the nuclear power plant.

6.47. Appropriate means shall be provided at the nuclear power plant for the monitoring of activity in fluid systems that have the potential for significant contamination, and for the collection of process samples.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants
Requirement 72: Compressed air systems
The design basis for any compressed air system that serves an item important to safety at the nuclear power plant shall specify the quality, flow rate and cleanness of the air to be provided.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants
Requirement 73: Air conditioning systems and ventilation systems
Systems for air conditioning, air heating, air cooling and ventilation shall be provided as appropriate in auxiliary rooms or other areas at the nuclear power plant to maintain the required environmental conditions for systems and components important to safety in all plant states.

6.48. Systems shall be provided for the ventilation of buildings at the nuclear power plant with appropriate capability for the cleaning of air:

To prevent unacceptable dispersion of airborne radioactive substances within the plant;

To reduce the concentration of airborne radioactive substances to levels compatible with the need for access by personnel to the area;

To keep the levels of airborne radioactive substances in the plant below authorized limits and as low as reasonably achievable;

To ventilate rooms containing inert gases or noxious gases without impairing the capability to controlradioactive effluents;

To control gaseous radioactivereleases to the environment below the authorized limits on discharges and to keep them as low as reasonably achievable.

6.49. Areas of higher contamination at the plant shall be maintained at a negative pressure differential (partial vacuum) with respect to areas of lower contamination and other accessible areas.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants
Requirement 74: Fire protection systems
Fire protection systems, including fire detection systems and fire extinguishing systems, fire containmentbarriers and smoke control systems, shall be provided throughout the nuclear power plant, with due account taken of the results of the fire hazardanalysis.

6.50. The fire protection systems installed at the nuclear power plant shall be capable of dealing safely with fire events of the various types that are postulated.

6.51. Fire extinguishing systems shall be capable of automatic actuation where appropriate. Fire extinguishing systems shall be designed and located to ensure that their rupture or spurious or inadvertent operation would not significantly impair the capability of items important to safety.

6.52. Fire detection systems shall be designed to provide operating personnel promptly with information on the location and spread of any fires that start.

6.53. Fire detection systems and fire extinguishing systems that are necessary to protect against a possible fire following a postulated initiating event shall be appropriately qualified to resist the effects of the postulated initiating event.

6.54. Non-combustible or fire retardant and heat resistant materials shall be used wherever practicable throughout the plant, in particular in locations such as the containment and the control room.

See also SSR-2/2 (Rev. 1) on Safety of Nuclear Power Plants: Commissioning and Operation Requirement 22, SSG-64on Protection against Internal Hazards in the Design of Nuclear Power Plants and SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants, as well as

SSG-77 , Protection Against Internal and External Hazards in the Operation of Nuclear Power Plants
Requirement 75: Lighting systems
Adequate lighting shall be provided in all operational areas of the nuclear power plant in operational states and in accident conditions.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants
Requirement 76: Overhead lifting equipment
Overhead lifting equipment shall be provided for lifting and lowering items important to safety at the nuclear power plant, and for lifting and lowering other items in the proximity of items important to safety.

6.55. The overhead lifting equipment shall be designed so that:

Measures are taken to prevent the lifting of excessive loads;

Conservative design measures are applied to prevent any unintentional dropping of loads that could affect items important to safety;

The plant layout permits safe movement of the overhead lifting equipment and of items being transported;

Such equipment can be used only in specified plant states (by means of safety interlocks on the crane);

Such equipment for use in areas where items important to safety are located is seismically qualified.

You may also be interested in:

SSG-62 Design of Auxiliary Systems and Supporting Systems for Nuclear Power Plants
Other power conversion systems Requirement 77: Steam supply system, feedwater system and turbine generators
The design of the steam supply system, feedwater system and turbine generators for the nuclear power plant shall be such as to ensure that the appropriate designlimits of the reactor coolant pressure boundary are not exceeded in operational states or in accident conditions.

6.56. The design of the steam supply system shall provide for appropriately rated and qualified steam isolation valves capable of closing under the specified conditions in operational states and in accident conditions.

6.57. The steam supply system and the feedwater systems shall be of sufficient capacity and shall be designed to prevent anticipated operational occurrences from escalating to accident conditions.

6.58. The turbine generators shall be provided with appropriate protection such as overspeed protection and vibration protection, and measures shall be taken to minimize the possible effects of turbine generated missiles on items important to safety.
Treatment of Radioactive Effluents and Radioactive Waste Requirement 78: Systems for treatment and control of waste
Systems shall be provided for treating solid radioactive waste and liquid radioactive waste at the nuclear power plant to keep the amounts and concentrations of radioactivereleases below the authorized limits on discharges and as low as reasonably achievable.

6.59. Systems and facilities shall be provided for the management and storage of radioactive waste on the nuclear power plant site for a period of time consistent with the availability of the relevant disposal option.

See also:

GSR Part 5 onPredisposal Management of Radioactive Waste

GSG-1 on Classification of Radioactive Waste

SSG-40 on Predisposal Management of Radioactive Waste

WS-G-6.1 on Storage of Radioactive Waste from Nuclear Power Plants and Research Reactors

6.60. The design of the plant shall incorporate appropriate features to facilitate the movement, transport and handling of radioactive waste. Consideration shall be given to the provision of access to facilities and to capabilities for lifting and for packaging.
Requirement 79: Systems for treatment and control of effluents
Systems shall be provided at the nuclear power plant for treating liquid and gaseous radioactive effluents to keep their amounts below the authorized limits on discharges and as low as reasonably achievable.

6.61. Liquid and gaseous radioactive effluents shall be treated at the plant so that exposure of members of the public due to discharges to the environment is as low as reasonably achievable.

6.62. The design of the plant shall incorporate suitable means to keep liquid radioactivereleases to the environmentas low as reasonably achievable and to ensure that radioactivereleases remain below the authorized limits on discharges.

6.63. The cleanup equipment for the gaseous radioactive substances shall provide the necessary retention factor to keep radioactivereleases below the authorized limits on discharges. Filter systems shall be designed so that their efficiency can be tested, their performance and function can be regularly monitored over their service life, and filter cartridges can be replaced while maintaining the throughput of air.

See also GSR Part 3 on Radiation Protection and Safety of Radiation Sources, International Basic Safety Standards, GSR Part 5 on Predisposal Management of Radioactive Waste, GSG-19 Monitoring for Protection of the Public and the Environment and GSG-10 Prospective Radiological Environmental Impact Assessment for Facilities and Activities
Fuel handling and storage systems Requirement 80: Fuel handling and storage systems
Fuel handling and storage systems shall be provided at the nuclear power plant to ensure that the integrity and properties of the fuel are maintained at all times during fuel handling and storage.

6.64. The design of the plant shall incorporate appropriate features to facilitate the lifting, movement and handling of fresh fuel and spent fuel.

See also SSG-63 on Design of Fuel Handling and Storage Systems for Nuclear Power Plants, SSG-73 on Core Management and Fuel Handling for Nuclear Power Plants, GSR Part 5 on Predisposal Management of Radioactive Waste, SSG-15 (Rev. 1) on Storage of Spent Nuclear Fuel, SSG-27 (Rev. 1) on Criticality Safety in the Handling of Fissile Material

6.65. The design of the plant shall be such as to prevent any significant damage to items important to safety during the transfer of fuel or casks, or in the event of fuel or casks being dropped.

6.66. The fuel handling and storage systems for irradiated and non-irradiated fuel shall be designed:

To prevent criticality by a specified margin, by physical means or by means of physical processes, and preferably by use of geometrically safe configurations, even under conditions of optimum moderation;

To permit inspection of the fuel;

To permit maintenance, periodic inspection and testing of components important to safety;

To prevent damage to the fuel;

To prevent the dropping of fuel in transit;

To provide for the identification of individual fuel assemblies;

To provide proper means for meeting the relevant requirements for radiation protection;

To ensure that adequate operating procedures and a system of accounting for, and control of, nuclear fuel can be implemented to prevent any loss of, or loss of control over, nuclear fuel.

6.67. In addition, the fuel handling and storage systems for irradiated fuel shall be designed:

To permit adequate removal of heat from the fuel in operational states and in accident conditions;

To prevent the dropping of spent fuel in transit;

To prevent causing unacceptable handling stresses on fuel elements or fuel assemblies;

To prevent the potentially damaging dropping of heavy objects such as spent fuel casks, cranes or other objects onto the fuel;

To permit safe keeping of suspect or damaged fuel elements or fuel assemblies;

To control levels of soluble absorber if this is used for criticalitysafety;

To facilitate maintenance and future decommissioning of fuel handling and storagefacilities;

To facilitate decontamination of fuel handling and storageareas and equipment when necessary;

To accommodate, with adequate margins, all the fuel removed from the reactor in accordance with the strategy for core management that is foreseen and the amount of fuel in the full reactor core;

To facilitate the removal of fuel from storage and its preparation for off-sitetransport.

6.68. For reactors using a water pool system for fuelstorage, the design shall be such as to prevent the uncovering of fuel assemblies in all plant states that are of relevance for the spent fuel pool so that the possibility of conditions arising that could lead to an early radioactive release or a large radioactive release is ‘practically eliminated’²⁶26 and so as to avoid high radiation fields on the site. The design of the plant:

shall provide the necessary fuel cooling capabilities;

shall provide features to prevent the uncovering of fuel assemblies in the event of a leak or a pipe break;

shall provide a capability to restore the water inventory.

The design shall also include features to enable the safe use of non-permanent equipment to ensure sufficient water inventory for the long term cooling of spent fuel and for providing shielding against radiation.²⁷27

6.68A. The design shall include the following:

Means for monitoring and controlling the water temperature for operational states and for accident conditions that are of relevance for the spent fuel pool;

Means for monitoring and controlling the water level for operational states and for accident conditions that are of relevance for the spent fuel pool;

Means for monitoring and controlling the activity in water and in air for operational states and means for monitoring the activity in water and in air for accident conditions that are of relevance for the spent fuel pool;

Means for monitoring and controlling the water chemistry for operational states.

Radiation protection Requirement 81: Design for radiation protection
Provision shall be made for ensuring that doses to operating personnel at the nuclear power plant will be maintained below the dose limits and will be kept as low as reasonably achievable, and that the relevant dose constraints will be taken into consideration.

6.69. Radiation sources throughout the plant shall be comprehensively identified, and exposures and radiation risks associated with them shall be kept as low as reasonably achievable²⁸28, the integrity of the fuelcladding shall be maintained, and the generation and transport of corrosion products and activation products shall be controlled.

See also GSR Part 3 on Radiation Protection and Safety of Radiation Sources, International Basic Safety Standards, GSG-7 Occupational Radiation Protection, SSG-90 on Radiation Protection Aspects of Design for Nuclear Power Plants

6.70. Materials used in the manufacture of structures, systems and components shall be selected to minimize activation of the material as far as is reasonably practicable.

6.71. For the purposes of radiation protection, provision shall be made for preventing the release or the dispersion of radioactive substances, radioactive waste and contamination at the plant.

6.72. The plant layout shall be such as to ensure that access of operating personnel to areas with radiationhazards and areas of possible contamination is adequately controlled, and that exposures and contamination are prevented or reduced by this means and by means of ventilation systems.

6.73. The plant shall be divided into zones that are related to their expected occupancy, and to radiation levels and contamination levels in operational states (including refuelling, maintenance and inspection) and to potential radiation levels and contamination levels in accident conditions. Shielding shall be provided so that radiation exposure is prevented or reduced.

6.74. The plant layout shall be such that the doses received by operating personnel during normal operation, refuelling, maintenance and inspection can be kept as low as reasonably achievable, and due account shall be taken of the necessity for any special equipment to be provided to meet these requirements.

6.75. Plant equipment subject to frequent maintenance or manual operation shall be located in areas of low dose rate to reduce the exposure of workers.

6.76. Facilities shall be provided for the decontamination of operating personnel and plant equipment.
Requirement 82: Means of radiation monitoring
Equipment shall be provided at the nuclear power plant to ensure that there is adequate radiationmonitoring in operational states and design basis accident conditions and, as far as is practicable, in design extension conditions.

6.77. Stationary dose rate meters shall be provided for monitoring local radiationdose rates at plant locations that are routinely accessible by operating personnel and where the changes in radiation levels in operational states could be such that access is allowed only for certain specified periods of time.

See also

GSR Part 3 on Radiation Protection and Safety of Radiation Sources, International Basic Safety Standards,

GSG-7 on Occupational Radiation Protection,

SSG-90 on Radiation Protection Aspects of Design for Nuclear Power Plants and

as well as

GSG-10 Prospective Radiological Environmental Impact Assessment for Facilities and Activities

6.78. Stationary dose rate meters shall be installed to indicate the general radiation levels at suitable plant locations in accident conditions. The stationary dose rate meters shall provide sufficient information in the control room or in the appropriate control position that operating personnel can initiate corrective actions if necessary.

6.79. Stationary monitors shall be provided for measuring the activity of radioactive substances in the atmosphere in those areas routinely occupied by operating personnel and where the levels of activity of airborne radioactive substances might be such as to necessitate protective measures. These systems shall provide an indication in the control room or in other appropriate locations when a high activity concentration of radionuclides is detected. Monitors shall also be provided in areas subject to possible contamination as a result of equipment failure or other unusual circumstances.

6.80. Stationary equipment and laboratory facilities shall be provided for determining, in a timely manner, the concentrations of selected radionuclides in fluid process systems, and in gas and liquid samples taken from plant systems or from the environment, in operational states and in accident conditions.

6.81. Stationary equipment shall be provided for monitoringradioactive effluents and effluents with possible contamination prior to or during discharges from the plant to the environment.

6.82. Instruments shall be provided for measuring surface contamination. Stationary monitors (e.g. portal radiation monitors, and hand and foot monitors) shall be provided at the main exit points from controlled areas and supervised areas to facilitate the monitoring of operating personnel and equipment.

6.83. Facilities shall be provided for monitoring for exposure and contamination of operating personnel. Processes shall be put in place for assessing and for recording the cumulative doses to workers over time.

6.84. Arrangements shall be made to assess exposures and other radiological impacts, if any, in the vicinity of the plant by environmental monitoring of dose rates or activity concentrations, with particular reference to:

Exposure pathways to people, including the food chain;

Radiological impacts, if any, on the local environment;

The possible buildup, and accumulation in the environment, of radioactive substances;

The possibility of there being any unauthorized routes for radioactivereleases.

References

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).

INTERNATIONAL ATOMIC ENERGY AGENCY, Safety Assessment for Facilities and Activities, IAEA Safety Standards Series No. GSR Part 4 (Rev. 1), IAEA, Vienna (2016).

INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safety Glossary: Terminology Used in Nuclear Safety and Radiation Protection (2007 Edition), IAEA, Vienna (2007).

INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Nuclear Power Plants: Commissioning and Operation, IAEA Safety Standards Series No. SSR-2/2 (Rev. 1), IAEA, Vienna (2016).

INTERNATIONAL NUCLEAR SAFETY ADVISORY GROUP, Defence in Depth in Nuclear Safety, INSAG-10, IAEA, Vienna (1996).

INTERNATIONAL NUCLEAR SAFETY ADVISORY GROUP, Basic Safety Principles for Nuclear Power Plants, 75-INSAG-3 Rev. 1, INSAG-12, IAEA, Vienna (1999).

INTERNATIONAL NUCLEAR SAFETY ADVISORY GROUP, Maintaining the Design Integrity of Nuclear Installations throughout their Operating Life, INSAG-19, IAEA, Vienna (2003).

INTERNATIONAL ATOMIC ENERGY AGENCY, The Management System for Facilities and Activities, IAEA Safety Standards Series No. GS-R-3, IAEA, Vienna (2006). (A revision of this publication is in preparation, to be issued as GSR Part 2).

EUROPEAN COMMISSION, 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 ENVIRONMENT PROGRAMME,WORLD HEALTH ORGANIZATION, Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards, IAEA Safety Standards Series No. GSR Part 3, IAEA, Vienna (2014).

INTERNATIONAL ATOMIC ENERGY AGENCY, Site Evaluation for Nuclear Installations, IAEA Safety Standards Series No. NS-R-3 (Rev. 1), IAEA, Vienna (2015).

FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL CIVIL AVIATION ORGANIZATION, INTERNATIONAL LABOUR ORGANIZATION, INTERNATIONAL MARITIME ORGANIZATION, INTERPOL, OECD NUCLEAR ENERGY AGENCY, PAN AMERICAN HEALTH ORGANIZATION, PREPARATORY COMMISSON FOR THE COMPREHENSIVE TEST-BAN TREATY ORGANIZATION, UNITED NATIONS ENVIRONMENT PROGRAMME, UNITED NATIONS OFFICE FOR THE CO-ORDINATION OF HUMANITARIAN AFFAIRS, WORLD HEALTH ORGANIZATION, WORLD METEOROLOGICAL ORGANIZATION, Preparedness and Response for a Nuclear or Radiological Emergency, IAEA Safety Standards Series No. GSR Part 7, IAEA, Vienna (2015).

[8] GS-R-3 has been revised and is now issued as GSR Part 2 Leadership and Management for Safety

[10] NS-R-3 (Rev. 1) has been revised and is now superseded by SSR-1 Site Evaluation for Nuclear Installations
8. Definitions
The following new and revised definitions differ from those in the IAEA Safety Glossary: Terminology Used in Nuclear Safety and Radiation Protection (2007 Edition), IAEA, Vienna (2007): http://www-pub.iaea.org/books/

The symbol (i) denotes an information note.

controlled state

Plant state, following an anticipated operational occurrence or accident conditions, in which the fundamental safety functions can be ensured and which can be maintained for a time sufficient to effect provisions to reach a safe state.

plant states (considered in design)

Operational states

Accident conditions

Normal operation

Anticipated operational occurences

Design basis accidents

Design extension conditions

Without significant fuel degradation

With core melting

accident conditions

Deviations from normal operation that are less frequent and more severe than anticipated operational occurrences.

(i) Accident conditions comprise design basis accidents and design extension conditions.

design basis accident

A postulated accident leading to accident conditions for which a facility is designed in accordance with established design criteria and conservative methodology, and for which releases of radioactive material are kept within acceptable limits.

design extension conditions

Postulated accident conditions that are not considered for design basis accidents, but that are considered in the design process for the facility in accordance with best estimate methodology, and for which releases of radioactive material are kept within acceptable limits.

(i) Design extension conditions comprise conditions in events without significant fuel degradation and conditions in events with core melting.

safe state

Plant state, following an anticipated operational occurrence or accident conditions, in which the reactor is subcritical and the fundamental safety functions can be ensured and maintained stable for a long time.

safety feature for design extension conditions

Item designed to perform a safety function or which has a safety function in design extension conditions.

safety system settings

Settings for levels at which safety systems are automatically actuated in the event of anticipated operational occurrences or design basis accidents, to prevent safety limits from being exceeded.

The 2018 Edition of the IAEA Safety Glossary is availabe here
Table Of Contents
Table Of Contents
Table Of Contents 1. Introduction
1. Introduction #
TOC \z \h 1. Introduction PAGE # Background
Background #
Background PAGE # Objective
Objective #
Objective PAGE # Scope
Scope #
Scope PAGE # Structure
Structure #
Structure PAGE # 2. Applying the safety principles and concepts
2. Applying the safety principles and concepts #
2. Applying the safety principles and concepts PAGE # Radiation protection in design
Radiation protection in design #
Radiation protection in design PAGE # Safety in design
Safety in design #
Safety in design PAGE # The concept of defence in depth
The concept of defence in depth #
The concept of defence in depth PAGE # Maintaining the integrity of design of the plant throughout the lifetime of the plant
Maintaining the integrity of design of the plant throughout the lifetime of the plant #
Maintaining the integrity of design of the plant throughout the lifetime of the plant PAGE # 3. Management of safety in design
3. Management of safety in design #
3. Management of safety in design PAGE # Requirement 1: Responsibilities in the management of safety in plant design
Requirement 1: Responsibilities in the management of safety in plant design #
Requirement 1: Responsibilities in the management of safety in plant design
PAGE # Requirement 2: Management system for plant design
Requirement 2: Management system for plant design #
Requirement 2: Management system for plant design PAGE # Requirement 3: Safety of the plant design throughout the lifetime of the plant
Requirement 3: Safety of the plant design throughout the lifetime of the plant #
Requirement 3: Safety of the plant design throughout the lifetime of the plant
PAGE # 4. Principal technical requirements
4. Principal technical requirements #
4. Principal technical requirements PAGE # Requirement 4: Fundamental safety functions
Requirement 4: Fundamental safety functions #
Requirement 4: Fundamental safety functions PAGE # Requirement 5: Radiation protection in design
Requirement 5: Radiation protection in design #
Requirement 5: Radiation protection in design PAGE # Requirement 6: Design for a nuclear power plant
Requirement 6: Design for a nuclear power plant #
Requirement 6: Design for a nuclear power plant PAGE # Requirement 7: Application of defence in depth
Requirement 7: Application of defence in depth #
Requirement 7: Application of defence in depth PAGE # Requirement 8: Interfaces of safety with security and safeguards
Requirement 8: Interfaces of safety with security and safeguards #
Requirement 8: Interfaces of safety with security and safeguards PAGE # Requirement 9: Proven engineering practices
Requirement 9: Proven engineering practices #
Requirement 9: Proven engineering practices PAGE # Requirement 10: Safety assessment
Requirement 10: Safety assessment #
Requirement 10: Safety assessment PAGE # Requirement 11: Provision for construction
Requirement 11: Provision for construction #
Requirement 11: Provision for construction PAGE # Requirement 12: Features to facilitate radioactive waste management and decommissioning
Requirement 12: Features to facilitate radioactive waste management and decommissioning #
Requirement 12: Features to facilitate radioactive waste management and decommissioning PAGE # 5. General plant design
5. General plant design #
5. General plant design PAGE # Design basis
Design basis #
Design basis PAGE # Requirement 13: Categories of plant states
Requirement 13: Categories of plant states #
Requirement 13: Categories of plant states PAGE # Requirement 14: Design basis for items important to safety
Requirement 14: Design basis for items important to safety #
Requirement 14: Design basis for items important to safety PAGE # Requirement 15: Design limits
Requirement 15: Design limits #
Requirement 15: Design limits PAGE # Requirement 16: Postulated initiating events
Requirement 16: Postulated initiating events #
Requirement 16: Postulated initiating events PAGE # Requirement 17: Internal and external hazards
Requirement 17: Internal and external hazards #
Requirement 17: Internal and external hazards PAGE # Internal hazards
Internal hazards #
Internal hazards PAGE # External hazards
External hazards #
External hazards PAGE # Requirement 18: Engineering design rules
Requirement 18: Engineering design rules #
Requirement 18: Engineering design rules PAGE # Requirement 19: Design basis accidents
Requirement 19: Design basis accidents #
Requirement 19: Design basis accidents PAGE # Requirement 20: Design extension conditions
Requirement 20: Design extension conditions #
Requirement 20: Design extension conditions PAGE # Combinations of events and failures
Combinations of events and failures #
Combinations of events and failures PAGE # Requirement 21: Physical separation and independence of safety systems
Requirement 21: Physical separation and independence of safety systems #
Requirement 21: Physical separation and independence of safety systems
PAGE # Requirement 22: Safety classification
Requirement 22: Safety classification #
Requirement 22: Safety classification PAGE # Requirement 23: Reliability of items important to safety
Requirement 23: Reliability of items important to safety #
Requirement 23: Reliability of items important to safety PAGE # Requirement 24: Common cause failures
Requirement 24: Common cause failures #
Requirement 24: Common cause failures PAGE # Requirement 25: Single failure criterion
Requirement 25: Single failure criterion #
Requirement 25: Single failure criterion PAGE # Requirement 26: Fail-safe design
Requirement 26: Fail-safe design #
Requirement 26: Fail-safe design PAGE # Requirement 27: Support service systems
Requirement 27: Support service systems #
Requirement 27: Support service systems PAGE # Requirement 28: Operational limits and conditions for safe operation
Requirement 28: Operational limits and conditions for safe operation #
Requirement 28: Operational limits and conditions for safe operation PAGE # Design for safe operation over the lifetime of the plant
Design for safe operation over the lifetime of the plant #
Design for safe operation over the lifetime of the plant PAGE # Requirement 29: Calibration, testing, maintenance, repair, replacement, inspection and monitoring of items important to safety
Requirement 29: Calibration, testing, maintenance, repair, replacement, inspection and monitoring of items important to safety #
Requirement 29: Calibration, testing, maintenance, repair, replacement, inspection and monitoring of items important to safety PAGE # Requirement 30: Qualification of items important to safety
Requirement 30: Qualification of items important to safety #
Requirement 30: Qualification of items important to safety PAGE # Requirement 31: Ageing management
Requirement 31: Ageing management #
Requirement 31: Ageing management PAGE # Human factors
Human factors #
Human factors PAGE # Requirement 32: Design for optimal operator performance
Requirement 32: Design for optimal operator performance #
Requirement 32: Design for optimal operator performance PAGE # Other design considerations
Other design considerations #
Other design considerations PAGE # Requirement 33: Safety systems, and safety features for design extension conditions, of units of a multiple unit nuclear power plant
Requirement 33: Safety systems, and safety features for design extension conditions, of units of a multiple unit nuclear power plant #
Requirement 33: Safety systems, and safety features for design extension conditions, of units of a multiple unit nuclear power plant PAGE # Requirement 34: Systems containing fissile material or radioactive material
Requirement 34: Systems containing fissile material or radioactive material #
Requirement 34: Systems containing fissile material or radioactive material
PAGE # Requirement 35: Nuclear power plants used for cogeneration of heat and power, heat generation or desalination
Requirement 35: Nuclear power plants used for cogeneration of heat and power, heat generation or desalination #
Requirement 35: Nuclear power plants used for cogeneration of heat and power, heat generation or desalination PAGE # Requirement 36: Escape routes from the plant
Requirement 36: Escape routes from the plant #
Requirement 36: Escape routes from the plant PAGE # Requirement 37: Communication systems at the plant
Requirement 37: Communication systems at the plant #
Requirement 37: Communication systems at the plant PAGE # Requirement 38: Control of access to the plant
Requirement 38: Control of access to the plant #
Requirement 38: Control of access to the plant PAGE # Requirement 39: Prevention of unauthorized access to, or interference with, items important to safety
Requirement 39: Prevention of unauthorized access to, or interference with, items important to safety #
Requirement 39: Prevention of unauthorized access to, or interference with, items important to safety PAGE # Requirement 40: Prevention of harmful interactions of systems important to safety
Requirement 40: Prevention of harmful interactions of systems important to safety #
Requirement 40: Prevention of harmful interactions of systems important to safety
PAGE # Requirement 41: Interactions between the electrical power grid and the plant
Requirement 41: Interactions between the electrical power grid and the plant #
Requirement 41: Interactions between the electrical power grid and the plant
PAGE # Safety analysis
Safety analysis #
Safety analysis PAGE # Requirement 42: Safety analysis of the plant design
Requirement 42: Safety analysis of the plant design #
Requirement 42: Safety analysis of the plant design PAGE # Deterministic approach
Deterministic approach #
Deterministic approach PAGE # Probabilistic approach
Probabilistic approach #
Probabilistic approach PAGE # 6. Design of specific plant systems
6. Design of specific plant systems #
6. Design of specific plant systems PAGE # Reactor core and associated features
Reactor core and associated features #
Reactor core and associated features PAGE # Requirement 43: Performance of fuel elements and assemblies
Requirement 43: Performance of fuel elements and assemblies #
Requirement 43: Performance of fuel elements and assemblies PAGE # Requirement 44: Structural capability of the reactor core
Requirement 44: Structural capability of the reactor core #
Requirement 44: Structural capability of the reactor core PAGE # Requirement 45: Control of the reactor core
Requirement 45: Control of the reactor core #
Requirement 45: Control of the reactor core PAGE # Requirement 46: Reactor shutdown
Requirement 46: Reactor shutdown #
Requirement 46: Reactor shutdown PAGE # Reactor coolant systems
Reactor coolant systems #
Reactor coolant systems PAGE # Requirement 47: Design of reactor coolant systems
Requirement 47: Design of reactor coolant systems #
Requirement 47: Design of reactor coolant systems PAGE # Requirement 48: Overpressure protection of the reactor coolant pressure boundary
Requirement 48: Overpressure protection of the reactor coolant pressure boundary #
Requirement 48: Overpressure protection of the reactor coolant pressure boundary
PAGE # Requirement 49: Inventory of reactor coolant
Requirement 49: Inventory of reactor coolant #
Requirement 49: Inventory of reactor coolant PAGE # Requirement 50: Cleanup of reactor coolant
Requirement 50: Cleanup of reactor coolant #
Requirement 50: Cleanup of reactor coolant PAGE # Requirement 51: Removal of residual heat from the reactor core
Requirement 51: Removal of residual heat from the reactor core #
Requirement 51: Removal of residual heat from the reactor core PAGE # Requirement 52: Emergency cooling of the reactor core
Requirement 52: Emergency cooling of the reactor core #
Requirement 52: Emergency cooling of the reactor core PAGE # Requirement 53: Heat transfer to an ultimate heat sink
Requirement 53: Heat transfer to an ultimate heat sink #
Requirement 53: Heat transfer to an ultimate heat sink PAGE # Containment structure and containment system
Containment structure and containment system #
Containment structure and containment system PAGE # Requirement 54: Containment system for the reactor
Requirement 54: Containment system for the reactor #
Requirement 54: Containment system for the reactor PAGE # Requirement 55: Control of radioactive releases from the containment
Requirement 55: Control of radioactive releases from the containment #
Requirement 55: Control of radioactive releases from the containment PAGE # Requirement 56: Isolation of the containment
Requirement 56: Isolation of the containment #
Requirement 56: Isolation of the containment PAGE # Requirement 57: Access to the containment
Requirement 57: Access to the containment #
Requirement 57: Access to the containment PAGE # Requirement 58: Control of containment conditions
Requirement 58: Control of containment conditions #
Requirement 58: Control of containment conditions PAGE # Instrumentation and control systems
Instrumentation and control systems #
Instrumentation and control systems PAGE # Requirement 59: Provision of instrumentation
Requirement 59: Provision of instrumentation #
Requirement 59: Provision of instrumentation PAGE # Requirement 60: Control systems
Requirement 60: Control systems #
Requirement 60: Control systems PAGE # Requirement 61: Protection system
Requirement 61: Protection system #
Requirement 61: Protection system PAGE # Requirement 62: Reliability and testability of instrumentation and control systems
Requirement 62: Reliability and testability of instrumentation and control systems #
Requirement 62: Reliability and testability of instrumentation and control systems
PAGE # Requirement 63: Use of computer based equipment in systems important to safety
Requirement 63: Use of computer based equipment in systems important to safety #
Requirement 63: Use of computer based equipment in systems important to safety
PAGE # Requirement 64: Separation of protection systems and control systems
Requirement 64: Separation of protection systems and control systems #
Requirement 64: Separation of protection systems and control systems PAGE # Requirement 65: Control room
Requirement 65: Control room #
Requirement 65: Control room PAGE # Requirement 66: Supplementary control room
Requirement 66: Supplementary control room #
Requirement 66: Supplementary control room PAGE # Requirement 67: Emergency response facilities on the site
Requirement 67: Emergency response facilities on the site #
Requirement 67: Emergency response facilities on the site PAGE # Emergency power supply
Emergency power supply #
Emergency power supply PAGE # Requirement 68: Design for withstanding the loss of off-site power
Requirement 68: Design for withstanding the loss of off-site power #
Requirement 68: Design for withstanding the loss of off-site power PAGE # Supporting systems and auxiliary systems
Supporting systems and auxiliary systems #
Supporting systems and auxiliary systems PAGE # Requirement 69: Performance of supporting systems and auxiliary systems
Requirement 69: Performance of supporting systems and auxiliary systems #
Requirement 69: Performance of supporting systems and auxiliary systems
PAGE # Requirement 70: Heat transport systems
Requirement 70: Heat transport systems #
Requirement 70: Heat transport systems PAGE # Requirement 71: Process sampling systems and post-accident sampling systems
Requirement 71: Process sampling systems and post-accident sampling systems #
Requirement 71: Process sampling systems and post-accident sampling systems
PAGE # Requirement 72: Compressed air systems
Requirement 72: Compressed air systems #
Requirement 72: Compressed air systems PAGE # Requirement 73: Air conditioning systems and ventilation systems
Requirement 73: Air conditioning systems and ventilation systems #
Requirement 73: Air conditioning systems and ventilation systems PAGE # Requirement 74: Fire protection systems
Requirement 74: Fire protection systems #
Requirement 74: Fire protection systems PAGE # Requirement 75: Lighting systems
Requirement 75: Lighting systems #
Requirement 75: Lighting systems PAGE # Requirement 76: Overhead lifting equipment
Requirement 76: Overhead lifting equipment #
Requirement 76: Overhead lifting equipment PAGE # Other power conversion systems
Other power conversion systems #
Other power conversion systems PAGE # Requirement 77: Steam supply system, feedwater system and turbine generators
Requirement 77: Steam supply system, feedwater system and turbine generators #
Requirement 77: Steam supply system, feedwater system and turbine generators
PAGE # Treatment of Radioactive Effluents and Radioactive Waste
Treatment of Radioactive Effluents and Radioactive Waste #
Treatment of Radioactive Effluents and Radioactive Waste PAGE # Requirement 78: Systems for treatment and control of waste
Requirement 78: Systems for treatment and control of waste #
Requirement 78: Systems for treatment and control of waste PAGE # Requirement 79: Systems for treatment and control of effluents
Requirement 79: Systems for treatment and control of effluents #
Requirement 79: Systems for treatment and control of effluents PAGE # Fuel handling and storage systems
Fuel handling and storage systems #
Fuel handling and storage systems PAGE # Requirement 80: Fuel handling and storage systems
Requirement 80: Fuel handling and storage systems #
Requirement 80: Fuel handling and storage systems PAGE # Radiation protection
Radiation protection #
Radiation protection PAGE # Requirement 81: Design for radiation protection
Requirement 81: Design for radiation protection #
Requirement 81: Design for radiation protection PAGE # Requirement 82: Means of radiation monitoring
Requirement 82: Means of radiation monitoring #
Requirement 82: Means of radiation monitoring PAGE # References
References #
References PAGE # 8. Definitions
8. Definitions #
8. Definitions PAGE #
1 ¹INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Nuclear Power Plants: Design, IAEA Safety Standards Series No. SSR-2/1, IAEA, Vienna (2012).

2 ²The possibility of certain conditions arising may be considered to have been ‘practically eliminated’ if it would be physically impossible for the conditions to arise or if these conditions could be considered with a high level of confidence to be extremely unlikely to arise.

3 ³An ‘early radioactive release’ in this context is a radioactiverelease for which off-siteprotective actions would be necessary but would be unlikely to be fully effective in due time. A ‘large radioactive release’ is a radioactiverelease for which off-siteprotective actions that are limited in terms of lengths of time and areas of application would be insufficient for the protection of people and of the environment.

4 ⁴The possibility of certain conditions arising may be considered to have been ‘practically eliminated’ if it would be physically impossible for the conditions to arise or if these conditions could be considered with a high level of confidence to be extremely unlikely to arise.

5 ⁵The design organization is the organization responsible for preparation of the final detailed design of the plant to be built.

6 ⁶ Requirements on the management system are established in IAEA Safety Standards Series No. GS-R-3 [8]. (A revision of this publication is in preparation, to be issued as GSR Part 2)

7 ⁷The possibility of certain conditions arising may be considered to have been ‘practically eliminated’ if it would be physically impossible for the conditions to arise or if these conditions could be considered with a high level of confidence to be extremely unlikely to arise.

8 ⁸ Requirements on radiationprotection and safety of radiation sources are established in IAEA Safety Standards Series No. GSR Part 3, RadiationProtection and Safety of Radiation Sources: International Basic Safety Standards [9].

9 ⁹A ‘cliff edge effect’, in a nuclear power plant, is an instance of severely abnormal plant behaviour caused by an abrupt transition from one plant status to another following a small deviation in a plant parameter, and thus a sudden large variation in plant conditions in response to a small variation in an input.

10 ¹⁰ Requirements on safety assessment for facilities and activities are established in GSR Part 4 (Rev. 1) [2].

11 ¹¹Requirements on site evaluation for nuclear installations are established in IAEA Safety Standards Series No. NS-R-3 (Rev. 1), Site Evaluation for Nuclear Installations [10].

12 ¹²A ‘cliff edge effect’, in a nuclear power plant, is an instance of severely abnormal plant behaviour caused by an abrupt transition from one plant status to another following a small deviation in a plant parameter, and thus a sudden large variation in plant conditions in response to a small variation in an input.

13 ¹³The analysis of design extension conditions for the plant could be performed by means of a best estimate approach (more stringent approaches may be used according to States’ requirements).

14 ¹⁴The possibility of certain conditions arising may be considered to have been ‘practically eliminated’ if it would be physically impossible for the conditions to arise or if these conditions could be considered with a high level of confidence to be extremely unlikely to arise.

15 ¹⁵ For returning the plant to a safe state or for mitigating the consequences of an accident, consideration could be given to the full design capabilities of the plant and to the temporary use of additional systems.

16 ¹⁶The possibility of certain conditions arising may be considered to have been ‘practically eliminated’ if it would be physically impossible for the conditions to arise or if these conditions could be considered with a high level of confidence to be extremely unlikely to arise.

17 ¹⁷A single failure is a failure that results in the loss of capability of a system or component to perform its intended safety function(s) and any consequential failure(s) that result from it. The single failure criterion is a criterion (or requirement) applied to a system such that it must be capable of performing its task in the presence of any single failure.

18 ¹⁸Requirements on safety assessment for facilities and activities are established in GSR Part 4 (Rev. 1) [2].

19 ¹⁹A ‘cliff edge effect’, in a nuclear power plant, is an instance of severely abnormal plant behaviour caused by an abrupt transition from one plant status to another following a small deviation in a plant parameter, and thus a sudden large variation in plant conditions in response to a small variation in an input.

20 ²⁰A ‘cliff edge effect’, in a nuclear power plant, is an instance of severely abnormal plant behaviour caused by an abrupt transition from one plant status to another following a small deviation in a plant parameter, and thus a sudden large variation in plant conditions in response to a small variation in an input.

21 ²¹In most cases, one containmentisolation valve or check valve is outside the containment and the other is inside the containment. Other arrangements might be acceptable, however, depending on the design.

22 ²²Non-permanent equipment need not necessarily be stored on the site.

23 ²³Emergency response facilities are addressed in IAEA Safety Standards Series No. GSR Part 7, Preparedness and Response for a Nuclear or Radiological Emergency [11]. For nuclear power plants, emergency response facilities (which are separate from the control room and the supplementary control room) include the technical support centre, the operational support centre and the emergency centre.

24 ²⁴A prime mover is a component (such as a motor, solenoid operator or pneumatic operator) that converts energy into action when commanded by an actuation device.

25 ²⁵Non-permanent equipment need not necessarily be stored on the site.

26 ²⁶The possibility of certain conditions arising may be considered to have been ‘practically eliminated’ if it would be physically impossible for the conditions to arise or if these conditions could be considered with a high level of confidence to be extremely unlikely to arise.

27 ²⁷Non-permanent equipment need not necessarily be stored on the site.

28 ²⁸ Requirements on radiation protection and the safety of radiation sources for facilities and activities are established in GSR Part 3 [9].

Definition

The achievement of proper operating conditions, prevention of accidents and mitigation of accident consequences, resulting in protection of workers, the public and the environment from undue radiation risks.

Scope notes

a. Often abbreviated to safety in IAEA publications on nuclear safety. Safety means nuclear safety unless otherwise stated, in particular when other types of safety (e.g. fire safety, conventional industrial safety) are also being discussed. See protection and safety for a discussion of the relationship between nuclear safety and radiation protection.

b. There is not an exact distinction between the general terms safety and security. In general, security is concerned with intentional actions by people that could cause or threaten harm to other people; safety is concerned with the broader issue of harmful consequences to people (and to the environment) arising from exposure to radiation, whatever the cause.

c. The interaction between arrangements for security and arrangements for safety depends on the context. Areas in which arrangements for safety and arrangements for security interact include, for example: the regulatory infrastructure; engineering provisions in the design and construction of nuclear installations and other facilities; controls on access to nuclear installations and other facilities; the categorization of radioactive sources; source design; the security of the management of radioactive sources and radioactive material; the recovery of sources that are not under regulatory control; emergency response plans; and radioactive waste management.

d. See protection and safety for a discussion of the relationship between nuclear safety and radiation protection.

Definition

That which is established or required by the Fundamental Safety Principles (IAEA Safety Fundamentals) [EUROPEAN ATOMIC ENERGY COMMUNITY, INTERNATIONAL ATOMIC ENERGY AGENCY, FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, 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).] or IAEA Safety Requirements publications or by (national or international) laws or regulations.

Scope notes

Caution: In IAEA publications, requirement (and ‘required’ and other words deriving from the verb ‘to require’) should be used in this sense only. Care should be taken to avoid confusion: the use of ‘requirement’ in the more general sense of something that is necessary should be avoided.

Requirements, including numbered ‘overarching’ requirements, are expressed as ‘shall’ statements. Reported (quoted) requirements, e.g. in a Safety Guide, are reported using a formulation such as ‘it is required to…’.

Definition

The protection of people from harmful effects of exposure to ionizing radiation, and the means for achieving this.

Scope notes

a. The International Commission on Radiological Protection and others use the term radiological protection, which is synonymous.

b. The accepted understanding of the term radiation protection is restricted to protection of people. Suggestions to extend the definition to include the protection of non-human species or the protection of the environment are controversial.

Definition

Any person who works, whether full time, part time or temporarily, for an employer and who has recognized rights and duties in relation to occupational radiation protection.

Scope notes

A self-employed person is regarded as having the duties of both an employer and a worker.

Definition

The conditions under which people, animals and plants live or develop and which sustain all life and development; especially such conditions as affected by human activities.

Scope notes

See also protection of the environment

Definition

For the purposes of radiation protection, radiation capable of producing ion pairs in biological material(s).

Definition

Strictly: relating to a nucleus; relating to or using energy released in nuclear fission or fusion. (adjective)

Scope notes

a. Caution: The adjective ‘nuclear’ is used in many phrases to modify a noun that it cannot logically modify. It must be borne in mind that the meaning of such phrases may be unclear (as opposed to nuclear).

b. Caution: The phrases may therefore need to be explained, and their usage may be open to misunderstanding, misrepresentation or mistranslation.

c. Caution: Such phrases include: nuclear accident; nuclear community; nuclear emergency; nuclear facility; nuclear fuel; nuclearincident; nuclear installation; nuclear material; nuclear medicine; (a) nuclear power; nuclear safety; and nuclear security.

Definition

1. A facility (including associated buildings and equipment) in which nuclear material is produced, processed, used, handled, stored or disposed of.

Scope notes

a. Also called a nuclear fuel cycle facility. See also facilities and activities and nuclear installation.

c. For safeguards purposes, see the definition of facility in the Safeguards Glossary [INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safeguards Glossary (2022 Edition), IAEA, Vienna].

b. A nuclear facility is an authorized facility, and this aspect of authorization is made explicit in other broadly similar definitions established in Nuclear Security Series publications, i.e. a nuclear facility is “A facility (including associated buildings and equipment) in which nuclear material is produced, processed, used, handled, stored or disposed of and for which an authorization or licence is required” [INTERNATIONAL ATOMIC ENERGY AGENCY, Nuclear Security Recommendations on Nuclear and Other Radioactive Material out of Regulatory Control, IAEA Nuclear Security Series No. 15, IAEA, Vienna (2011)] or “A facility (including associated buildings and equipment) in which nuclear material is produced, processed, used, handled, stored or disposed of and for which a specific licence is required” [INTERNATIONAL ATOMIC ENERGY AGENCY, Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities, INFCIRC/225/Revision 5, IAEA Nuclear Security Series No. 13, IAEA, Vienna (2011)].

<h2>Definition</h2>1. A multiattribute quantity expressing hazard, danger or chance of harmful or injurious consequences associated with exposures or potential exposures. It relates to quantities such as the probability that specific deleterious consequences may arise and the magnitude and character of such consequences.<h2>Scope notes</h2><ul><li>In mathematical terms, this can be expressed generally as a set of triplets, <img src="/PoolParty/OpenConfigPathResourceProvider/frontendroot?fileName=custom/iaea-safety-glossary/images/docimg/formula-risk.jpg" style= "vertical-align: middle" style= "vertical-align: middle" alt="formula" height="30">, where Sis an identification or description of a scenario i, pis the probability of that scenario and Xis a measure of the consequence of the scenario. The concept of risk is sometimes also considered to include uncertainty in the probabilities pof the scenarios.</li></ul>

Definition

1. The process and the result of developing a concept, detailed plans, supporting calculations and specifications for a facility and its parts.

Scope notes

The terms siting, design, construction, commissioning, operation and decommissioning are normally used to delineate the six major stages of the lifetime of an authorized facility and of the associated licensing process. In the special case of disposal facilities for radioactive waste, decommissioning is replaced in this sequence by closure.

Definition

1. Any unintended event, including operating errors, equipment failures and other mishaps, the consequences or potential consequences of which are not negligible from the point of view of protection and safety.

Scope notes

Caution: In the 2008 INES Manual (INTERNATIONAL ATOMIC ENERGY AGENCY, INES: The International Nuclear and Radiological Event Scale User’s Manual, 2008 Edition, IAEA, Vienna (2013)), there was a fundamental mismatch between the terminology used in safety standards and the designations used in INES. In short, events that would be considered accidents according to the safety standards definition may be accidents or incidents (i.e. not accidents) in INES terminology. This was not a serious day to day problem because the two areas are quite separate and have quite different purposes. However, it was a potential cause of confusion in communication with the news media and the public.

Definition

Loss of the ability of a structure, system or component to function within acceptance criteria.

Scope notes

a. Caution: Note that the structure, system or component is considered to fail when it becomes incapable of functioning, whether or not this is needed at that time.

b. Caution: A failure in, for example, a backup system may not be manifest until the system is called upon to function, either during testing or on failure of the system it is backing up.

c. A failure may be the result of, for example, a hardware fault, a software fault, a system fault, an operator error or a maintenance error.

Definition

Accident more severe than a design basis accident and involving significant core degradation.

Definition

A postulated or assumed set of conditions and/or events.

Scope notes

See event.

Most commonly used in analysis or assessment to represent possible future conditions and/or events to be modelled, such as possible accidents at a nuclear facility, or the possible future evolution of a disposal facility and its surroundings. A scenario may represent the conditions at a single point in time or a single event, or a time history of conditions and/or events (including processes)

Definition

1. The function or power or (usually as controls) means of directing, regulating or restraining.

Scope notes

It should be noted that the usual meaning of the English word control in safety related contexts is somewhat ‘stronger’ (more active) than that of its usual translations and other similar words in some other languages. For example, control typically implies not only checking or monitoring something but also ensuring that corrective or enforcement measures are taken if the results of the checking or monitoring indicate such a need. This is in contrast, for example, to the more limited usage of the equivalent word in French and Spanish.

Definition

2. Any material that is either special fissionable material or source material as defined in Article XX of the IAEA Statute.

Scope notes

a. Nuclear material is necessary for the production of nuclear weapons or other nuclear explosive devices. Under comprehensive safeguards agreements, the IAEA verifies that all nuclear material subject to safeguards has been declared and placed under safeguards.

b. Certain non-nuclear materials are essential for the use or production of nuclear material and may also be subject to IAEA safeguards under certain agreements.

d. The Statute of the IAEA [Statute of the International Atomic Energy Agency, IAEA, Vienna (1990).] uses the term special fissionable material, with the meaning essentially of nuclear material as defined here, but explicitly excluding source material.

c. For safeguards purposes, see the definition of nuclear material in the Safeguards Glossary [INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safeguards Glossary (2022 Edition), IAEA, Vienna].

Definition

1. The prevention and detection of, and response to, criminal or intentional unauthorized acts involving nuclear material, other radioactive material, associated facilities or associated activities.

Scope notes

a. Often abbreviated to security in IAEA publications on nuclear security.

b. Security of nuclear material for reasons relating to non-proliferation of nuclear weapons is outside the scope of the IAEA safety standards and of the IAEA Nuclear Security Series.

e. The interaction between arrangements for security and arrangements for safety depends on the context. Areas in which arrangements for safety and arrangements for security interact include, for example: the regulatory infrastructure; engineering provisions in the design and construction of nuclear installations and other facilities; controls on access to nuclear installations and other facilities; the categorization of radioactive sources; source design; the security of the management of radioactive sources and radioactive material; the recovery of sources that are not under regulatory control; emergency response plans; and radioactive waste management.

d. There is not an exact distinction between the general terms safety and security. In general, security is concerned with intentional actions by people that could cause or threaten harm to other people; safety is concerned with the broader issue of harmful consequences to people (and to the environment) arising from exposure to radiation, whatever the cause.

c. Security of nuclear material for reasons relating to non-proliferation of nuclear weapons is outside the scope of the IAEA safety standards and of the IAEA Nuclear Security Series.

Definition

Any action that might be taken, condition that might be applied or procedure that might be followed to fulfil the requirements of Safety Requirements.

Definition

All activities performed to achieve the purpose for which an authorized facility was constructed.

Scope notes

c. See also abnormal operation and normal operation.

a. For a nuclear power plant, this includes maintenance, refuelling, in-service inspection and other associated activities.

b. The terms siting, design, construction, commissioning, operation and decommissioning are normally used to delineate the six major stages of the lifetime of an authorized facility and of the associated licensing process. In the special case of disposal facilities for radioactive waste, decommissioning is replaced in this sequence by closure.

Definition

The process of manufacturing and assembling the components of a facility, the carrying out of civil works, the installation of components and equipment, and the performance of associated tests.

Scope notes

The terms siting, design, construction, commissioning, operation and decommissioning are normally used to delineate the six major stages of the lifetime of an authorized facility and of the associated licensing process. In the special case of disposal facilities for radioactive waste, decommissioning is replaced in this sequence by closure.

Definition

1. An authority or a system of authorities designated by the government of a State as having legal authority for conducting the regulatory process, including issuing authorizations, and thereby regulating the nuclear, radiation, radioactive waste and transportsafety.

Scope notes

a. The regulatory body is generally a national entity, established and empowered by law, whose organization, management, functions, processes, responsibilities and competences are subject to the requirements of IAEA safety standards.

b. The national competent authority for the regulation of radioactive materialtransport (see INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport of Radioactive Material — 2018 Edition, IAEA Safety Standards Series No. SSR-6 (Rev. 1), IAEA, Vienna (2018)) is included in this description, as is the regulatory body for protection and safety.

c. Caution: Supersedes the term regulatory authority, which should not be used.

Definition

Evaluation of the potential hazards associated with the operation of a facility or the conduct of an activity.

Scope notes

b. Safety analysis is often used interchangeably with safety assessment. However, when the distinction is important, safety analysis should be used as a documented process for the study of safety, and safety assessment should be used as a documented process for the evaluation of safety — for example, evaluation of the magnitude of hazards, evaluation of the performance of safety measures and judgement of their adequacy, or quantification of the overall radiological impact or safety of a facility or activity.

a. The formal safety analysis is part of the overall safety assessment; that is, it is part of the systematic process that is carried out throughout the design process (and throughout the lifetime of the facility or the activity) to ensure that all the relevant safety requirements are met by the proposed (or actual) design.

Definition

A systematic reassessment of the safety of an existing facility (or activity) carried out at regular intervals to deal with the cumulative effects of ageing, modifications, operating experience, technical developments and siting aspects, and aimed at ensuring a high level of safety throughout the service life of the facility (or activity).

Definition

A general term encompassing all of the elements (items) of a facility or activity that contribute to protection and safety, except human factors.

Scope notes

Human factors may be reflected in structures, systems and components in so far as ergonomics — the study of people’s efficiency in their work setting — is an element in their design.

<h2>Definition</h2>In the context of the reporting and analysis of events, an event is any occurrence unintended by the operator, including operating error, equipment failure or other mishap, and deliberate action on the part of others, the consequences or potential consequences of which are not negligible from the point of view of protection and safety.<h2>Scope notes</h2><ul><li>c. This difference derives from the fact that event reporting and analysis is concerned directly with the question of whether an event that could develop into an accident with significant consequences actually does so; terms such as accident are used only to describe the end result and therefore other terms are needed to describe the earlier stages. See also initiating event and initiating event: postulated initiating event.</li><li>a. Caution: As with INES, the terminology related to the reporting and analysis of events is not consistent with the terminology used in safety standards, and great care should be taken to avoid confusion.</li><li>b. Caution: In particular, the definition of event given above is identical in essence to the safety standards definition (1) of accident.</li><li>d. Event is also used in the phrase ‘features, events and processes’ associated with the site and the facility in the context of site characterization for a disposal facility for radioactive waste.</li><li>e. Relevant features, events and processes relating to the site are those that might influence the long term performance of the disposal facility and thus could affect safety. These are addressed in a safety case and in a supporting safety assessment. Types of events and circumstances <img src="https://nucleus.iaea.org/sites/nss-oui/Figures/i_8c833feb-4169-44c2-969e-a4337573ef03.JPG " style="width: 480px;"></li></ul>

Definition

The organized activity, both administrative and technical, of keeping structures, systems and components in good operating condition, including both preventive and corrective (or repair) aspects.

Definition

1. Administrative and technical actions taken to allow the removal of some or all of the regulatory controls from a facility.

Scope notes

Note: The terms siting, design, construction, commissioning, operation and decommissioning are normally used to delineate the six major stages of the lifetime of an authorized facility and of the associated licensing process. In the special case of disposal facilities for radioactive waste, decommissioning is replaced in this sequence by closure.

a. Caution: This does not apply for that part of a disposal facility in which radioactive waste is emplaced, or for certain facilities used for the disposal of naturally occurring radioactive material (NORM) or of residues from the mining and processing of radioactive ores. For all of these the term closure is used instead of decommissioning.

b. Caution: Decommissioning typically includes dismantling of the facility (or part thereof) to reduce the associated radiation risks, but in the IAEA’s usage this need not be the case. A facility could, for example, be decommissioned without dismantling and the existing structures subsequently put to another use (after decontamination).

c. The use of the term decommissioning implies that no further use of the facility (or part thereof) for its existing purpose is foreseen.

d. Actions for decommissioning are taken at the end of the operating lifetime of a facility to retire it from service with due regard for the health and safety of workers and members of the public and protection of the environment.

e. Subject to national legal and regulatory requirements, a facility (or its remaining parts) may also be considered decommissioned if it is incorporated into a new or existing facility, even if the site on which it is located is still under regulatory control or institutional control.

f. The actions will need to be such as to ensure the long term protection of the public and protection of the environment, and typically include reducing the levels of residual radionuclides in the materials and on the site of the facility so that the materials can be safely recycled, reused or disposed of as exempt waste or as radioactive waste and the site can be released for unrestricted use or otherwise reused.

g. Decommissioning can entail activities that are similar to remediation (also an authorized process), such as removal of contaminated soil from an area within the authorized boundary of a facility, but in this case, such removals are normally referred to as cleanup activities and are typically performed under the authorization for decommissioning.

Definition

1. The process of determining whether the quality or performance of a product or service is as stated, as intended or as required.

Scope notes

Verification is closely related to quality management and quality control.

Definition

A specific purpose that must be accomplished for safety for a facility or activity to prevent or to mitigate radiological consequences of normal operation, anticipated operational occurrences and accident conditions. [INTERNATIONAL ATOMIC ENERGY AGENCY, Safety Classification of Structures, Systems and Components in Nuclear Power Plants, IAEA Safety Standards Series No. SSG-30, IAEA, Vienna (2014).]

Scope notes

[INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Nuclear Power Plants: Design, IAEA Safety Standards Series No. SSR 2/1 (Rev. 1), IAEA, Vienna (2016).] establishes requirements on safety functions to be fulfilled by the design of a nuclear power plant in order to meet three general safety requirements:
(a) The capability to safely shut down the reactor and maintain it in a safe shutdown condition during and after appropriate operational states and accident conditions;
(b) The capability to remove residual heat from the reactor core, the reactor and nuclear fuel in storage after shutdown, and during and after appropriate operational states and accident conditions; (c) The capability to reduce the potential for the release of radioactive material and to ensure that any releases are within [prescribed limit]s during and after operational states and within acceptable limits during and after design basis accidents
This guidance is commonly condensed into a succinct expression of three fundamental safety functions for nuclear power plants:
(a) Control of reactivity;
(b) Cooling of radioactive material;
(c) Confinement of radioactive material.

<h2>Definition</h2>For nuclear power plants: <ul><li>(a) Control of reactivity; </li><li>(b) Cooling of radioactive material; </li><li>(c) Confinement of radioactive material. </li></ul><h2>Scope notes</h2><ul><li></ul>In some IAEA publications, ‘basic safety function’ and ‘main safety function’ are also used.</li></ul>

<h2>Definition</h2>1. A hierarchical deployment of different levels of diverse equipment and procedures to prevent the escalation of anticipated operational occurrences and to maintain the effectiveness of physical barriers placed between a radiation source or radioactive material and workers, members of the public or the environment, in operational states and, for some barriers, in accident conditions.<h2>Scope notes</h2><ul><li>a. The objectives of defence in depth are: <ul><li>(a) To compensate for human induced events and component failures; </li><li>(b) To maintain the effectiveness of the barriers by averting damage to the facility and to the barriers themselves; </li><li>(c) To protect workers, members of the public and the environment from harm in accident conditions in the event that these barriers are not fully effective</li></ul></li><li>c. Five levels of defence in depth are discussed in SSR-2/1 (Rev. 1) (See [INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Nuclear Power Plants: Design, IAEA Safety Standards Series No. SSR 2/1 (Rev. 1), IAEA, Vienna (2016).] for further information): <ul><li>(a) The purpose of the first level of defence is to prevent deviations from normal operation and the failure of items important to safety. </li><li>(b) The purpose of the second level of defence is to detect and control deviations from normal operation in order to prevent anticipated operational occurrences from escalating to accident conditions. </li><li>(c) The purpose of the third level of defence is to prevent damage to the reactor core and releases of radioactive material requiring off-site protective actions and to return the plant to a safe state by means of inherent and/or engineered safety features, safety systems and procedures. </li><li>(d) The purpose of the fourth level of defence is to prevent the progress of, and to mitigate the consequences of, accidents that result from failure of the third level of defence by preventing accident sequences that lead to large release of radioactive material or early release of radioactive material from occurring. </li><li>(e) The purpose of the fifth and final level of defence is to mitigate radiological consequences of a large release of radioactive material or an early release of radioactive material that could potentially result from an accident</li></ul></li><li>d. The International Nuclear Safety Group (INSAG) defined five levels of defence in depth (See [INTERNATIONAL NUCLEAR SAFETY ADVISORY GROUP, Defence in Depth in Nuclear Safety, INSAG Series No. 10, IAEA, Vienna (1996).] for further information): <ul><li>(a) Level 1: Prevention of abnormal operation and failures. </li><li>(b) Level 2: Control of abnormal operation and detection of failures. </li><li>(c) Level 3: Control of accidents within the design basis. </li><li>(d) Level 4: Control of severe plant conditions, including prevention of accident progression and mitigation of the consequences of severe accidents. </li><li>(e) Level 5: Mitigation of radiological consequences of significant releases of radioactive material.</li><li>e. The levels of defence are sometimes grouped into three safety layers: hardware, software and management control.</li><li>f. In the context of waste disposal, the term multiple barriers is used to describe a similar concept.</li><li>g. Note that defence in depth is used with a different meaning in the IAEA Nuclear Security Series in the context of nuclear security.</li><li>b. The Fundamental Safety Principles (IAEA Safety Fundamentals) [EUROPEAN ATOMIC ENERGY COMMUNITY, INTERNATIONAL ATOMIC ENERGY AGENCY, FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, 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).] (para. 3.31) states that: “Defence in depth is implemented primarily through the combination of a number of consecutive and independent levels of protection that would have to fail before harmful effects could be caused to people or to the environment. If one level of protection or barrier were to fail, the subsequent level or barrier would be available. When properly implemented, defence in depth ensures that no single technical, human or organizational failure could lead to harmful effects, and that the combinations of failures that could give rise to significant harmful effects are of very low probability. The independent effectiveness of the different levels of defence is a necessary element of defence in depth”.</li></ul>

Definition

Caution: When used in IAEA publications, the term radiation usually refers to ionizing radiation only. The IAEA has no statutory responsibilities in relation to non-ionizing radiation.

Scope notes

Ionizing radiation can be divided into low linear energy transfer radiation and high linear energy transfer radiation (as a guide to its relative biological effectiveness), or into strongly penetrating radiation and weakly penetrating radiation (as an indication of its ability to penetrate shielding or the human body).

Definition

4. radiation risks: Detrimental health effects of exposure to radiation (including the likelihood of such effects occurring), and any other safety related risks (including those to the environment) that might arise as a direct consequence of:
- Exposure to radiation;
- The presence of radioactive material (including radioactive waste) or its release to the environment;
- A loss of control over a nuclear reactor core, nuclear chain reaction, radioactive source or any other source of radiation. (From Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, IAEA, Vienna (2006).].)

Scope notes

For the purposes of the IAEA safety standards, it is assumed that there is no threshold level of radiation dose below which there are no associated radiation risks.

Safety Requirements and Safety Guides specify the radiation exposures and other radiation risks to which they refer.

Definition

The process of determining what level of protection and safety would result in the magnitude of individual doses, the number of individuals (workers and members of the public) subject to exposure and the likelihood of exposure being as low as reasonably achievable (ALARA), economic and social factors being taken into account.

Scope notes

Caution: The acronym ALARA should not be used to mean optimization of protection and safety.

See optimization of protection and safety.

Definition

Detrimental health effects of exposure to radiation (including the likelihood of such effects occurring), and any other safety related risks (including those to the environment) that might arise as a direct consequence of:
- (a) Exposure to radiation;
- (b) The presence of radioactive material (including radioactive waste) or its release to the environment;
- (c) A loss of control over a nuclear reactor core, nuclear chain reaction, radioactive source or any other source of radiation. (From [EUROPEAN ATOMIC ENERGY COMMUNITY, INTERNATIONAL ATOMIC ENERGY AGENCY, FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, 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).].)

Scope notes

For the purposes of the IAEA safety standards, it is assumed that there is no threshold level of radiation dose below which there are no associated radiation risks.

Safety Requirements and Safety Guides specify the radiation exposures and other radiation risks to which they refer.

Definition

A general term encompassing nuclear facilities, uses of all sources of ionizing radiation, all radioactive waste management activities, transport of radioactive material and other practices or circumstances in which people may be subject to exposure to radiation from naturally occurring or artificial sources.

Scope notes

a. The term facilities and activities is intended to provide an alternative to the terminology of sources and practices (or interventions) to refer to general categories of situations.

b. For example, a practice may involve many different facilities and/or activities, whereas the general definition (1) of source is too broad in some cases: a facility or activity might constitute a source, or might involve the use of many sources, depending upon the interpretation used.

c. The term facilities and activities is very general, and includes those for which little or no regulatory control may be necessary or achievable: the more specific terms authorized facility and authorized activity should be used to distinguish those facilities and activities for which any form of authorization has been given.

d. In the Fundamental Safety Principles (IAEA Safety Fundamentals), the term ‘facilities and activities — existing and new — utilized for peaceful purposes’ is abbreviated for convenience to facilities and activities as a general term encompassing any human activity that may cause people to be exposed to radiation risks arising from naturally occurring or artificial sources (see EUROPEAN ATOMIC ENERGY COMMUNITY, INTERNATIONAL ATOMIC ENERGY AGENCY, FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, 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), para. 1.9).

f. Facilities and activities are listed as follows in [INTERNATIONAL ATOMIC ENERGY AGENCY, Safety Assessment for Facilities and Activities, IAEA Safety Standards Series No. GSR Part 4 (Rev. 1), IAEA, Vienna (2015)]:
‘Facilities’ includes:
(a) Nuclear power plants;
(b) Other reactors (such as research reactors and critical assemblies);
(c) Enrichment facilities and nuclear fuel fabrication facilities;
(d) Conversion facilities used to generate uranium hexafluoride (UF₆);
(e) Storage facilities and reprocessing plants for irradiated fuel;
(f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of;
(g) Any other places where radioactive materials are produced, processed, used, handled or stored;
(h) Irradiation installations for medical, industrial, research and other purposes, and any places where radiation generators are installed;
(i)Facilities where the mining and processing of radioactive ores (such as ores of uranium and thorium) are carried out
‘Activities’ includes:
(a) The production, use, import and export of radiation sources for medical, industrial, research and other purposes;
(b) The transport of radioactive material;
(c) The decommissioning of facilities and the closure of repositories for radioactive waste;
(d) The close-out of facilities where the mining and processing of radioactive ores was carried out;
(e) Activities for radioactive waste management such as the discharge of effluents;
(f) The remediation of sites affected by residues from past activities

e. For safeguards purposes, see the definition of facility in the Safeguards Glossary [IAEA Safeguards Glossary (2022 Edition), IAEA, Vienna].

Definition

These include: nuclear facilities; irradiation installations; some mining and raw material processing facilities such as uranium mines; radioactive waste management facilities; and any other places where radioactive material is produced, processed, used, handled, stored or disposed of — or where radiation generators are installed — on such a scale that consideration of protection and safety is required.

Scope notes

‘Facilities’ includes:
(a) Nuclear power plants;
(b) Other reactors (such as research reactors and critical assemblies);
(c) Enrichment facilities and nuclear fuel fabrication facilities;
(d) Conversion facilities used to generate uranium hexafluoride (UF₆);
(e) Storage facilities and reprocessing plants for irradiated fuel;
(f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of;
(g) Any other places where radioactive materials are produced, processed, used, handled or stored;
(h) Irradiation installations for medical, industrial, research and other purposes, and any places where radiation generators are installed;
(i)Facilities where the mining and processing of radioactive ores (such as ores of uranium and thorium) are carried out

Definition

2. These include: the production, use, import and export of radiation sources for industrial, research and medical purposes; the transport of radioactive material; the decommissioning of facilities; radioactive waste management activities such as the discharge of effluents; and some aspects of the remediation of sites affected by residues from past activities.

Scope notes

The intention is to include any human activity that introduces additional sources of radiation or additional exposure pathways, or that modifies the network of exposure pathways from existing sources, so as to increase the exposure or the likelihood of exposure of people or the number of people exposed.

‘Activities’ includes:
(a) The production, use, import and export of radiation sources for medical, industrial, research and other purposes;
(b) The transport of radioactive material;
(c) The decommissioning of facilities and the closure of repositories for radioactive waste;
(d) The close-out of facilities where the mining and processing of radioactive ores was carried out;
(e) Activities for radioactive waste management such as the discharge of effluents;
(f) The remediation of sites affected by residues from past activities

Definition

1. The state or condition of being subject to irradiation.

Scope notes

Caution: Exposure should not be used as a synonym for dose. Dose is a measure of the effects of exposure.

Exposure to ionizing radiation can be broadly divided into categories of exposure (see categories of exposure) according to the status of the individual(s) exposed; into exposure situations according to the circumstances of the exposure; and according to the source of the exposure.

Definition

2. Designated in national law or by a regulatory body as being subject to regulatory control because of its radioactivity. (adjective)

Scope notes

Caution: This is the ‘regulatory’ definition, and should not be confused with the ‘scientific’ definition (1).

Definition

The action or process of setting free or being set free, or of allowing or being allowed to move or flow freely.

Scope notes

Caution: Release is used in both a physical ‘scientific’ sense (see discharge (1)) and a ‘regulatory’ sense (see clearance), as well as in the usual sense of, for example, a release of energy.

Definition

States defined under normal operation and anticipated operational occurrences.

Scope notes

Some States and organizations use the term operating conditions (in contrast to accident conditions) for this concept.

Definition

1. A source containing radioactive material that is used as a source of radiation.

Definition

Fissionable nuclear material in the form of fabricated elements for loading into the reactor core of a civil nuclear power plant or research reactor.

Definition

1. For legal and regulatory purposes, waste that contains, or is contaminated with, radionuclides at activity concentrations greater than clearance levels as established by the regulatory body.

Scope notes

a. In effect, radioactive material in gaseous, liquid or solid form for which no further use is foreseen.

b. Caution: It should be recognized that this definition is purely for regulatory purposes, and that material with activity concentrations equal to or less than clearance levels is radioactive from a physical viewpoint, although the associated radiological hazards are considered negligible. See radioactive, radioactive material, radioactive substance.

c. Waste should be used in the singular unless reference is expressly being made to the presence of various types of waste.

Definition

1. Anything that may cause radiation exposure — such as by emitting ionizing radiation or by releasing radioactive substances or radioactive material — and can be treated as a single entity for purposes of protection and safety.

Scope notes

b. A complex or multiple installation situated at one location or site may, as appropriate, be considered a single source for the purposes of application of safety standards.

a. For example, materials emitting radon are sources in the environment; a sterilization gamma irradiation unit is a source for the practice of irradiation preservation of food and sterilization of other products; an X ray unit may be a source for the practice of radiodiagnosis; a nuclear power plant is part of the practice of generating electricity by nuclear fission, and may be regarded as a source (e.g. with respect to discharges to the environment) or as a collection of sources (e.g. for purposes of occupational radiation protection purposes).

Definition

1. The period during which an authorized facility is used for its intended purpose, until decommissioning or closure.

Scope notes

The synonyms operating period and operational period are also used.

Definition

The process of selecting a suitable site for a facility, including appropriate assessment and definition of the related design bases.

Scope notes

The siting process for a nuclear installation generally consists of site survey and site selection.

The terms siting, design, construction, commissioning, operation and decommissioning are normally used to delineate the six major stages of the lifetime of an authorized facility and of the associated licensing process. In the special case of disposal facilities for radioactive waste, decommissioning is replaced in this sequence by closure.

Definition

The process by means of which systems and components of facilities and activities, having been constructed, are made operational and verified to be in accordance with the design and to have met the required performance criteria.

Scope notes

Commissioning may include both non-nuclear and/or non-radioactive and nuclear and/or radioactive testing.

The terms siting, design, construction, commissioning, operation and decommissioning are normally used to delineate the six major stages of the lifetime of an authorized facility and of the associated licensing process. In the special case of disposal facilities for radioactive waste, decommissioning is replaced in this sequence by closure.

Definition

1. The deliberate physical movement of radioactive material (other than that forming part of the means of propulsion) from one place to another.

Scope notes

a. The term transportation is also used, in particular in US English or where there is a need to distinguish this meaning of transport from meaning (2).

c. In some Nuclear Security Series publications addressing transport of nuclear material, transport has also been defined as “International or domestic carriage of nuclear material by any means of transport, beginning with the departure from a nuclear facility of the shipper and ending with the arrival at a nuclear facility of the receiver” [Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities, INFCIRC/225/Revision 5, IAEA Nuclear Security Series No. 13, IAEA, Vienna (2011)].

b. In the context of the Transport Regulations, transport comprises all operations and conditions associated with, and involved in, the movement of radioactive material; these include the design, manufacture, maintenance and repair of packaging, and the preparation, consigning, loading, carriage including in-transit storage, shipment after storage, unloading and receipt at the final destination of loads of radioactive material and packages.

Definition

1. Material designated in national law or by a regulatory body as being subject to regulatory control because of its radioactivity.

Scope notes

a. Some publications in the Nuclear Security Series extend the definition, for the avoidance of doubt, to indicate that “In the absence of such a designation by a State, any material for which protection is required by the current version of the International Basic Safety Standards” [Objective and Essential Elements of a State’s Nuclear Security Regime, IAEA Nuclear Security Series No. 20, IAEA, Vienna (2013)].

b. Caution: This is the ‘regulatory’ meaning of radioactive (2), and should not be confused with the ‘scientific’ meaning of radioactive (1): ‘exhibiting radioactivity; emitting or relating to the emission of ionizing radiation or particles’.

c. Caution: The ‘scientific’ meaning of radioactive (1) — as in radioactive substance — refers only to the presence of radioactivity, and gives no indication of the magnitude of the hazard involved.

d. Caution: The term radioactive substance is also used to indicate that the ‘scientific’ meaning of radioactive (see radioactive (1)) is intended, rather than the ‘regulatory’ meaning of radioactive (see radioactive (2)) suggested by the term radioactive material.

e. Caution: However, in some States the term radioactive substance is used for the ‘regulatory’ purpose. It is therefore essential that any such distinctions in meaning are clarified.

f.Caution: In some Nuclear Security Series publications (e.g. Refs [Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities, INFCIRC/225/Revision 5, IAEA Nuclear Security Series No. 13, IAEA, Vienna (2011), Nuclear Security Recommendations on Radioactive Material and Associated Facilities, IAEA Nuclear Security Series No. 14, IAEA, Vienna (2011)]) it is stated or implied that radioactive substance is synonymous with radioactive material, in order to confirm consistency with the 2005 CPPNM Amendment [Amendment to the Convention on the Physical Protection of Nuclear Material, INFCIRC/274/Rev. 1/Mod. 1 (Corrected), IAEA, Vienna (2021).]. In contexts related to the CPPNM Amendment, the terms are synonymous, but in general they may have different meanings.

g. In regulatory terminology in some States, radioactive material ceases to be radioactive material when it becomes radioactive waste; the term radioactive substance is used to cover both, that is radioactive substance includes radioactive material and radioactive waste.

h. Caution:Radioactive material should be used in the singular unless reference is expressly being made to the presence of various types of radioactive material.

Definition

2. Absorbed dose, committed equivalent dose, committed effective dose, equivalent dose, effective dose or organ dose, as indicated by the context.

Definition

The value of the effective dose or the equivalent dose to individuals in planned exposure situations that is not to be exceeded.

Definition

The value of a quantity used in certain specified activities or circumstances that must not be exceeded.

Scope notes

a. Caution: The term limit should only be used for a criterion that must not be exceeded; for example, where exceeding the limit would cause some form of legal sanction to be invoked.

b. Caution: Criteria used for other purposes — for example, to indicate a need for closer investigation or a review of procedures, or as a threshold for reporting to a regulatory body — should be described using other terms, such as reference level.

Definition

3. The systematic process that is carried out throughout the design process (and throughout the lifetime of the facility or the activity) to ensure that all the relevant safety requirements are met by the proposed (or actual) design.

Scope notes

Stages in the lifetime of a facility or activity at which a safety assessment is carried out and updated and the results are used by the designers, the operating organization and the regulatory body include:
(a) Site evaluation for the facility or activity;
(b) Development of the design;
(c) Construction of the facility or implementation of the activity;
(d) Commissioning of the facility or of the activity;
(e) Commencement of operation of the facility or conduct of the activity;
(f) Normal operation of the facility or normal conduct of the activity;
(g) Modification of the design or operation;
(h) Periodic safety reviews;
(i) Life extension of the facility beyond its original design life;
(j) Changes in ownership or management of the facility;
(k) Decommissioning of a facility;
(l) Closure of a disposal facility for radioactive waste and the post-closure phase;
(m) Remediation of a site and release from regulatory control
See INTERNATIONAL ATOMIC ENERGY AGENCY, Safety Assessment for Facilities and Activities, IAEA Safety Standards Series No. GSR Part 4 (Rev. 1), IAEA, Vienna (2016).

Safety assessment includes, but is not limited to, the formal safety analysis; that is, it includes the evaluation of the potential hazards associated with the operation of a facility or the conduct of an activity.

Definition

Operation within specified operational limits and conditions.

Scope notes

For a nuclear power plant, this includes startup, power operation, shutting down, shutdown, maintenance, testing and refuelling.

Definition

A deviation of an operational process from normal operation that is expected to occur at least once during the operating lifetime of a facility but which, in view of appropriate design provisions, does not cause any significant damage to items important to safety or lead to accident conditions.

Scope notes

Examples of anticipated operational occurrences are loss of normal electrical power and faults such as a turbine trip, malfunction of individual items of a normally running plant, failure to function of individual items of control equipment, and loss of power to the main coolant pump.

Some States and organizations use the term abnormal operation (for contrast with normal operation) for this concept.

Definition

Deviations from normal operation that are less frequent and more severe than anticipated operational occurrences.

Scope notes

Examples of such deviations include a major fuelfailure or a loss of coolant accident (LOCA).

See accident, event.

Accident conditions are considered in the design and comprise design basis accidents and design extension conditions.

Definition

A postulated event identified in design as capable of leading to anticipated operational occurrences or accident conditions.

Scope notes

The primary causes of postulated initiating events may be credible equipment failures and operator errors (both within and external to the facility), human induced events or natural events.

Definition

Item that is designed to perform a safety function for or that has a safety function for design extension conditions.

Scope notes

The concept of safety features for design extension conditions also applies for research reactors and nuclear fuel cycle facilities.

Definition

A system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents.

Scope notes

a. Safety systems consist of the protection system, the safety actuation systems and the safety system support features.

b. Components of safety systems may be provided solely to perform safety functions, or may perform safety functions in some plant operational states and non-safety functions in other operational states.

Definition

The taking of a set of actions during the evolution of an accident:
(a) To prevent the escalation to a severe accident;
(b) To mitigate the consequences of a severe accident;
(c) To achieve a long term safe stable state

Scope notes

b. By extension, accident management for a severe accident includes the taking of a set of actions during the evolution of the accident to mitigate degradation of the reactor core.

a. Aspect (b) of accident management (to mitigate the consequences of a severe accident) is also termed severe accident management.

Definition

A series of specified actions conducted in a certain order or manner.

Scope notes

The set of actions to be taken to conduct an activity or to perform a process is typically specified in a set of instructions.

Definition

2. The organization (and its contractors) which undertakes the siting, design, construction, commissioning and/or operation of a nuclear facility.

Scope notes

a. Caution: This usage is particular to documentation relating to the safety of radioactive waste management, with the corresponding understanding of siting as a multistage process.

b. Caution: This difference is partly a reflection of the particularly crucial role of siting in the safety of repositories.

Definition

Outside the site area.

Definition

1. An action for the purposes of avoiding or reducing doses that might otherwise be received in an emergency exposure situation or an existing exposure situation.

Scope notes

This is related to radiation protection (see protection (1), and protection and safety).

Definition

A release of radioactive material for which off-siteprotective actions that are limited in terms of times and areas of application are insufficient for protecting people and the environment.

Scope notes

See also early release of radioactive material; see also defence in depth.

Definition

The concept of practical elimination applies to plant event sequences that could lead to unacceptable consequences (i.e. an early radioactive release or a large radioactive release) that cannot be mitigated by reasonably practicable means. Practical elimination implies that those plant event sequences have to be demonstrated to be either physically impossible or, with a high level of confidence, extremely unlikely to arise by implementing safety provisions in the form of design and operational features.

Scope notes

b. Caution: The phrase ‘practically eliminated’ is misleading as it actually concerns the possible exclusion of event sequences from hypothetical scenarios rather than practicalities of safety. The phrase can also all too readily be misinterpreted, misrepresented or mistranslated as referring to the ‘elimination’ of ‘accidents’ by practical measures (or else ‘practically’ in the sense of ‘almost’?). Clear drafting in natural language would be preferable.

a. Practical elimination is part of a general approach to design safety and complements the adequate implementation of the concept of defence in depth.

c. Practical elimination can be achieved by, for example, the implementation of safety provisions in the form of design and operational features, either to ensure that plant event sequences that could lead to an early radioactive release or a large radioactive release are physically impossible or to demonstrate, with a high level of confidence, that such plant event sequences are extremely unlikely to arise (Preliminary definition from The IAEA Nuclear Safety and Security Glossary 2022 before revision in SSG-88).

Definition

The cessation of operation of a facility.

Definition

A physical obstruction that prevents or inhibits the movement of people, radionuclides or some other phenomenon (e.g. fire), or provides shielding against radiation.

Definition

An item that is part of a safety group and/or whose malfunction or failure could lead to radiationexposure of the site personnel or members of the public.

Scope notes

Items important to safety include:
- Those structures, systems and components whose malfunction or failure could lead to undue radiationexposure of site personnel or members of the public;
- Those structures, systems and components that prevent anticipated operational occurrences from leading to accident conditions;
- safety features for design extension conditions;
- Those features that are provided to mitigate the consequences of malfunction or failure of structures, systems and components.

Definition

Part of quality management intended to verify that structures, systems and components correspond to predetermined requirements.

Scope notes

This definition is taken from ISO 921:1997 (Nuclear Energy: Vocabulary) [INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, Nuclear Energy: Vocabulary (Second Edition), ISO 921:1997, ISO, Geneva (1997).]. A more general definition of quality control and definitions of related terms can be found in ISO 8402:1994 [INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, Quality Management and Quality Assurance — Vocabulary, ISO 8402:1994, ISO, Geneva (1994).].

Definition

One of the parts that make up a system.

Scope notes

a. A component may be a hardware component (e.g. wires, transistors, integrated circuits, motors, relays, solenoids, pipes, fittings, pumps, tanks and valves) or a software component (e.g. modules, routines, programmes, software functions).

b. A component may be made up of other components.

d. See also active component, passive component and core components.

c. The terms ‘equipment’, ‘component’ and ‘module’ are often used interchangeably. The relationship of these terms is not yet standardized.

Definition

The potential for harm or other detriment, especially for radiation risks; a factor or condition that might operate against safety.

Definition

2. A set of interrelated or interacting activities that transforms inputs into outputs.

Scope notes

A product is the result or output of a process.

Definition

Inspection of structures, systems and components undertaken over the operating lifetime by or on behalf of the operating organization for the purpose of identifying age related degradation or conditions that, if not addressed, might lead to the failure of structures, systems or components.

Scope notes

Inspection of operational activities, processes, etc., by or on behalf of the operating organization would normally be described by terms such as self-assessment and audit.

Definition

A departure from specified requirements.

Definition

An identified event that leads to anticipated operational occurrences or accident conditions.

Scope notes

a. This term (often shortened to initiator) is used in relation to event reporting and analysis, i.e. when such events have occurred.

b. For the consideration of hypothetical events at the design stage, the term postulated initiating event is used.

Definition

Plant state, following an anticipated operational occurrence or accident conditions, in which the reactor is subcritical and the fundamental safety functions can be ensured and maintained stable for a long time.

Definition

1. Radioactive substances on surfaces, or within solids, liquids or gases (including the human body), where their presence is unintended or undesirable, or the process giving rise to their presence in such places.

Scope notes

a. Also used less formally to refer to a quantity, namely the activity on a surface (or on a unit area of a surface).

c. Caution: The term contamination may have a connotation that is not intended. The term contamination refers only to the presence of radioactivity, and gives no indication of the magnitude of the hazard involved.

b. Contamination does not include residual material remaining at a site after the completion of decommissioning.

d. Caution:The term ‘contaminated evidence’ connotes a different meaning for the forensic scientist and for the nuclear forensic scientist. In general forensic science usage, ‘contaminated evidence’ refers to evidence contaminated with extraneous material, and thus compromised. Care should be taken to ensure that there is no possibility of confusion in referring to the presence of radionuclides on or within physical evidence: the term ‘evidence contaminated with radionuclides’ should be used to clarify this meaning.

Definition

A release of radioactive material for which off-siteprotective actions are necessary but are unlikely to be fully effective in due time.

Scope notes

See also large release of radioactive material and defence in depth.

Definition

A facility or location necessary for supporting an emergency response, for which specific functions are to be assigned at the preparedness stage, and which need to be usable under emergency conditions.

Scope notes

b. For both types, advance preparations are necessary to ensure their operability under emergency conditions. Depending on the emergency preparedness category and on the nature of an emergency, an emergency response facility may be designated an emergency response location.

c. For a nuclear power plant and other facilities in emergency preparedness category 1, emergency response facilities (which are separate from the control room and the supplementary control room) include: the technical support centre, from which technical support can be provided to the operating personnelin the control room in an emergency; the operational support centre, from which operational control can be maintained by personnel performing tasks at or near the facility; and the emergency centre, from which the on-site emergency response is managed.

a. There are two different types of emergency response facility or location: those established in advance (e.g. a technical support centre for a nuclear power plant) and those designated at the time of an emergency (e.g. a medical screening and triage area).

Definition

A description of the objectives, policy and concept of operations for the response to an emergency and of the structure, authorities and responsibilities for a systematic, coordinated and effective response. The emergency plan serves as the basis for the development of other plans, procedures and checklists.

Scope notes

a. Emergency plans are prepared at several different levels: international, national, regional, local and facility. They may include all activities planned to be carried out by all relevant organizations and authorities, or may be primarily concerned with the actions to be carried out by a particular organization. The term overall emergency plan is sometimes used for clarification when the former meaning is intended.

b. Details regarding the accomplishment of specific tasks outlined in an emergency plan are contained in emergency procedures.

Definition

A set of instructions describing in detail the actions to be taken by emergency workers in an emergency.

Definition

Within the site area.

Definition

The performance of actions to mitigate the consequences of an emergency for human life and health, property and the environment.

Scope notes

The emergency response also provides a basis for the resumption of normal social and economic activity.

Definition

The amount and isotopic composition of radioactive material released (or postulated to be released) from a facility.

Scope notes

Used in modelling releases of radionuclides to the environment, in particular in the context of accidents at nuclear installations or releases from radioactive waste in repositories.

Definition

A design, contracting or manufacturing organization supplying a service, component or facility.

Definition

Any person or organization to whom a registrant or licensee assigns duties, totally or partially, in relation to the design, manufacture, production or construction of a source.

Scope notes

An importer of a source is considered a supplier of the source.

The term supplier (of a source) includes designers, manufacturers, producers, constructors, assemblers, installers, distributors, sellers, importers or exporters of a source.

Definition

Any organization, corporation, partnership, firm, association, trust, estate, public or private institution, group, political or administrative entity, or other persons designated in accordance with national legislation who or which has responsibility and authority for any action having implications for protection and safety.

Scope notes

Supersedes the term [legal person], which is contrasted in legal texts with ‘natural person’, meaning an individual.

Definition

The granting by a regulatory body or other governmental body of written permission for a person or organization (the operator) to conduct specified activities.

Scope notes

a. Authorization could include, for example, licensing (issuing a licence), certification (issuing a certificate) or registration.

b. The term authorization is also sometimes used to describe the document granting such permission.

c. In nuclear security, the governmental body concerned will be a competent authority, and the definition is set out in Nuclear Security Series publications as “The granting by a competent authority of written permission for operation of an associated facility or for carrying out an associated activity [or a document granting such permission]” [Nuclear Security Recommendations on Radioactive Material and Associated Facilities, IAEA Nuclear Security Series No. 14, IAEA, Vienna (2011), Nuclear Security Recommendations on Nuclear and Other Radioactive Material out of Regulatory Control, IAEA Nuclear Security Series No. 15, IAEA, Vienna (2011), Objective and Essential Elements of a State’s Nuclear Security Regime, IAEA Nuclear Security Series No. 20, IAEA, Vienna (2013)].

d. Authorization is generally a more formal process than approval. In context of the Transport Regulations [INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport of Radioactive Material — 2018 Edition, IAEA Safety Standards Series No. SSR-6 (Rev. 1), IAEA, Vienna (2018).] approval is specific type of authorization concerning the items listed in para. 802 of the Transport Regulations that is carried out in accordance with the applicable requirements of the Transport Regulations.

e. See also approval: multilateral approval and unilateral approval.

Definition

Any person or organization applying to a regulatory body for authorization (or approval) to undertake specified activities.

Scope notes

Strictly, an applicant would be such from the time at which an application is submitted until the requested authorization is either granted or refused. However, the term is often used a little more loosely than this, in particular in cases where the authorization process is long and complex.

Definition

A passive elements (e.g. buildings, vessels, shielding).

Definition

A set of interrelated or interacting elements (system) for establishing policies and objectives and enabling the objectives to be achieved in an efficient and effective manner.

Scope notes

a. The component parts of the management system include the organizational structure, resources and organizational processes.

b. Management is defined (in ISO 9000) [INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, Quality Management Systems — Fundamentals and Vocabulary, ISO 9000:2015, ISO, Geneva (2015).] as coordinated activities to direct and control an organization.

c. The management system integrates all elements of an organization into one coherent system to enable all of the organization’s objectives to be achieved. These elements include the organizational structure, resources and processes.

d. Personnel, equipment and organizational culture as well as the documented policies and processes form parts of the management system.

e. The organization’s processes have to address the totality of the requirements on the organization as established in, for example, IAEA safety standards and other international codes and standards.

Definition

1. Any organization or person applying for authorization or authorized to operate an authorized facility or to conduct an authorized activity and responsible for its safety.

Scope notes

a. Caution: Note that such an organization may be the operating organization before operation starts.

c. Operating organization includes either those who are directly in control of a facility or an activity during use of a source (such as radiographers or carriers) or, in the case of a source not under control (such as a lost or illicitly removed source or a re-entering satellite), those who were responsible for the source before control over it was lost.

b. This includes, inter alia, private individuals, governmental bodies, consignors or carriers, licensees, hospitals and self-employed persons.

d. In practice, for an authorized facility, the operating organization is normally also the registrant or licensee. However, the separate terms are retained to refer to the two different capacities. See also operator

Definition

1. A legal document issued by the regulatory body granting authorization to perform specified activities relating to a facility or activity.

Scope notes

b. Authorization may take other forms, such as registration or certification.

a. A licence is a product of the authorization process (although the term licensing process is sometimes used), and a practice with a current licence is an authorized practice.

Definition

The range of conditions and events taken explicitly into account in the design of structures, systems and components and equipment of a facility, according to established criteria, such that the facility can withstand them without exceeding authorized limits.

Scope notes

Used as a noun, with the definition above. Also often used as an adjective, applied to specific categories of conditions or events to mean ‘included in the design basis’; as, for example, in design basis accident, design basis external events and design basis earthquake.

Definition

2. Confirmation by examination and by means of objective evidence that specified objectives have been met and specified requirements for specific results have been fulfilled. See validation.

Scope notes

a. The corresponding status is termed ‘verified’.

b. Verification typically entails the assessment of the results of an individual activity against its inputs.

c. Verification may comprise activities such as: performing alternative calculations; comparing a new design specification with a similar proven design specification; undertaking tests and demonstrations; and reviewing documents prior to issue.

Definition

2. Confirmation by examination and by means of objective evidence that specified objectives have been met and specified requirements for a specific intended purpose and use or application have been fulfilled. See also verification.

Scope notes

a. The corresponding status is termed ‘validated’.

b. Validation typically entails the assessment of a final product against its specified objectives and specified requirements.

c. The conditions of use for validation purposes may be real or simulated.

Definition

The granting of consent by a regulatory body.

Scope notes

Typically used to represent any form of consent from the regulatory body that does not meet the definition of authorization. In context of the Transport Regulations [INTERNATIONAL ATOMIC ENERGY AGENCY, Regulations for the Safe Transport of Radioactive Material — 2018 Edition, IAEA Safety Standards Series No. SSR-6 (Rev. 1), IAEA, Vienna (2018)] approval is a specific type of authorization concerning the items listed in para. 802 of the Transport Regulations that is carried out in accordance with the applicable requirements of the Transport Regulations.

Definition

The probability that a system or component or an item will meet its minimum performance requirements when called upon to do so, for a specified period of time and under stated operating conditions.

Scope notes

See also availability.

The reliability of a computer based system, for example, includes the reliability of hardware, which is usually quantified, and the reliability of software, which is usually a qualitative measure as there are no generally recognized means of quantifying the reliability of software.

Definition

1. The process of determining what level of protection and safety would result in the magnitude of individual doses, the number of individuals (workers and members of the public) subject to exposure and the likelihood of exposure being as low as reasonably achievable, economic and social factors being taken into account (ALARA).

Definition

The assembly of characteristics and attitudes in organizations and individuals which establishes that, as an overriding priority, protection and safety issues receive the attention warranted by their significance.

Scope notes

For a more detailed discussion, see [INTERNATIONAL NUCLEAR SAFETY ADVISORY GROUP, Safety Culture, Safety Series No. 75-INSAG-4, IAEA, Vienna (1992).].

Definition

The process of identifying and documenting the characteristics of a facility’s structures, systems and components (including computer systems and software), and of ensuring that changes to these characteristics are properly developed, assessed, approved, issued, implemented, verified, recorded and incorporated into the facility documentation.

Scope notes

‘Configuration’ is used in the sense of the physical, functional and operational characteristics of the structures, systems and components and parts of a facility.

<h2>Definition</h2>plant states (considered in design) <img src="https://vocabulary.iaea.org/https://nucleus.iaea.org/sites/nss-oui/Figures/i_dd7e8ea8-daac-40b0-b095-12bc565d655e.JPG" style="width: 480px;"> <h2>Scope notes</h2><ul><li>a. Caution: The entries that follow (terms and definitions) relate to consideration at the design stage (i.e. by means of hypothetical scenarios)</li><li>b. Caution: Care needs to be taken to select, use and relate defined terms and other words in such a way that clear distinctions are drawn and may be inferred between, for example: events and situations (see the entry for event); accidents and other incidents; what is actual (i.e. what is), possible (i.e. what might be) or potential (i.e. what could become), and what is hypothetical (i.e. what is postulated or assumed); and what is observed or determined objectively, and what is decided or declared subjectively.</li><li>c. Caution: ‘Conditions’, for example, is used in terms in the sense of rules set in design (as in operational limits and conditions) and also circumstances of operation (as in plant conditions); and in terms used in both design and operation (e.g. in accident conditions, service conditions).</li><li>f. See also event, model, probabilistic safety assessment, uncertainty.</li><li>e. The concept of facility states as it is used in the safety standards for research reactors and for nuclear fuel cycle facilities is broadly equivalent to the concept of plant states for nuclear power plants. Unless otherwise indicated, the definitions of terms grouped under ‘plant states’ apply for nuclear power plants, research reactors and nuclear fuel cycle facilities.</li><li>d. Caution: Drafters and reviewers thus need to bear in mind whether text concerns design or operation, or both. The potential, the postulated or the assumed in design needs to be distinguished from the observed or the determined in operation; and the decided on or declared (such as an emergency), in both design and operation, needs to be distinguished from the former (i.e. the potential, the postulated, the assumed, the observed and the determined).</li></ul>

Definition

For a nuclear chain reacting medium:
r = 1 - 1 / k_eff>
where k_eff is the ratio between the number of fissions in two succeeding generations (later to earlier) of the chain reaction.

Scope notes

A measure of the deviation from criticality of a nuclear chain reacting medium, such that positive values correspond to a supercritical state and negative values correspond to a subcritical state.

Definition

Prevention or control of releases of radioactive material to the environment in operation or in accidents.

Scope notes

a. Confinement is closely related in meaning to containment, but confinement is typically used to refer to the safety function of preventing the ‘escape’ of radioactive material, whereas containment refers to the means for achieving that function.

b. Caution: The Transport Regulations adopt a different distinction between confinement and containment, namely that confinement relates to preventing criticality and containment to preventing releases of radioactive material (see confinement system and containment system).

c. The main issue here is the differences in usage between the safety of nuclear installations and safety in the transport of radioactive material. Both terms, containment and confinement, are used in both areas (in the Transport Regulations, in the form of confinement system and containment system), and the usages of containment are (it seems) conceptually consistent, but the usages of confinement are not. Confinement in nuclear safety is the safety function that is performed by the containment.

Definition

2. Continuous or periodic measurement of radiological or other parameters or determination of the status of a structure, system or component.

Scope notes

a. Sampling may be involved as a preliminary step to measurement.

b. Although the concept is not fundamentally different from definition (1), this definition is more suited to the types of monitoring concerned primarily with safety (i.e. keeping sources under control) rather than with protection (i.e. controlling exposure).

c. This definition is particularly relevant to monitoring the status of a nuclear installation by tracking plant variables, or monitoring the long term performance of a waste disposal facility by tracking variables such as water fluxes.

d. These examples differ from definition (1) in that the routine measurements are themselves of no particular interest; the monitoring is only intended to detect unexpected deviations if they occur.

Definition

The protection of people against exposure to ionizing radiation or exposure due to radioactive material and the safety of sources, including the means for achieving this, and the means for preventing accidents and for mitigating the consequences of accidents if they do occur.

Scope notes

a. Safety is primarily concerned with maintaining control over sources, whereas (radiation) protection is primarily concerned with controlling exposure to radiation and its effects.

b. Clearly the two are closely connected: radiation protection (or radiological protection) is very much simpler if the source in question is under control, so safety necessarily contributes towards protection.

c. Sources come in many different types, and hence safety may be termed the safety of nuclear installations, radiation safety, the safety of radioactive waste management or safety in the transport of radioactive material, but protection (in this sense) is primarily concerned with protecting people against exposure, whatever the source, and so is always radiation protection.

d. For the purposes of the IAEA safety standards, protection and safety includes the protection of people against ionizing radiation and safety; it does not include non-radiation-related aspects of safety.

e. Protection and safety is concerned with both radiation risks under normal circumstances and radiation risks as a consequence of incidents, as well as with other possible direct consequences of a loss of control over a nuclear reactor core, nuclear chain reaction, radioactive source or any other source of radiation.

f. Safety measures include actions to prevent incidents and arrangements put in place to mitigate their consequences if they were to occur.

Definition

Standards issued pursuant to Article III(A)(6)⁸ of the Statute of the IAEA [Statute of the International Atomic Energy Agency, IAEA, Vienna (1990).].

⁸ “[The Agency is authorized…] 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 (including such standards for labour conditions)…”

Scope notes

a. Requirements, regulations, standards, rules, codes of practice or recommendations established to protect people and the environment against ionizing radiation and to minimize danger to life and property.

b. Safety standards issued since 1997 in the IAEA Safety Standards Series are designated as Safety Fundamentals, Safety Requirements or Safety Guides.

c. Some safety standards issued prior to 1997 in the (defunct) Safety Series were designated Safety Standards, Codes, Regulations or Rules.

d. Furthermore, some publications issued in the (defunct) Safety Series were not safety standards, notably those designated Safety Practices or Procedures and Data.

e. Other IAEA publications, such as Safety Reports and TECDOCs (most of which are issued pursuant to Article VIII of the Statute), are not safety standards.

Definition

A set of rules setting forth parameter limits, the functional capability and the performance levels of equipment and personnel approved by the regulatory body for safe operation of an authorized facility.

Definition

The period of time during which a facility or component is expected to perform according to the technical specifications to which it was produced.

Definition

1. Planned and controlled release of (usually gaseous or liquid) radioactive substances to the environment.

Scope notes

Strictly, the act or process of releasing the radioactive substances, but also used to describe the radioactive substances released.

Definition

An instance of severely abnormal conditions caused by an abrupt transition from one status of a facility to another following a small deviation in a parameter or a small variation in an input value.

Scope notes

In a nuclear power plant or nuclear fuel cycle facility, a cliff edge effect is an instance of severely abnormal facility behaviour caused by an abrupt transition from one facility status to another following a small deviation in a facility parameter, and thus a sudden large variation in facility conditions in response to a small variation in an input.

Definition

Any person or organization applying for authorization or authorized and/or responsible for safety when undertaking activities or in relation to any nuclear facilities or sources of ionizing radiation.

Scope notes

a. In some publications in the Nuclear Security Series, this term has been also defined as “Any person, organization, or government entity licensed or authorized to undertake the operation of an associated facility or to perform an associated activity” [Nuclear Security Recommendations on Physical Protection of Nuclear Material and Nuclear Facilities, INFCIRC/225/Revision 5, IAEA Nuclear Security Series No. 13, IAEA, Vienna (2011)].

e. Synonymous with operating organization.

d. Operator includes either those who are directly in control of a facility or an activity during use or transport of a source (such as radiographers or carriers) or, in the case of a source not under control (such as a lost or illicitly removed source or a re-entering satellite), those who were responsible for the source before control over it was lost.

c. Caution: Operator is sometimes used to refer to operating personnel (e.g. control room operators). If used in this way, particular care should be taken to ensure that there is no possibility of confusion.

b. Operator includes, inter alia, private individuals, governmental bodies, consignors or carriers, licensees, hospitals, self-employed persons.

Definition

A comprehensive, structured approach to identifying failure scenarios, constituting a conceptual and mathematical tool for deriving numerical estimates of risk.

Scope notes

Three levels of probabilistic safety assessment are generally recognized:
- Level 1 comprises the assessment of failures leading to determination of the frequency of fuel damage.
- Level 2 includes the assessment of containment response, leading, together with Level 1 results, to the determination of frequencies of failure of the containment and release to the environment of a given percentage of the reactor core’s inventory of radionuclides.
- Level 3 includes the assessment of off-site consequences, leading, together with the results of Level 2 analysis, to estimates of public risks
(See, for example, [INTERNATIONAL NUCLEAR SAFETY ADVISORY GROUP, Probabilistic Safety Assessment, Safety Series No. 75-INSAG-6, IAEA, Vienna (1994).].)

<h2>Definition</h2>1. All administrative and operational activities involved in the handling, pretreatment, treatment, conditioning, transport, storage and disposal of radioactive waste <img src="https://vocabulary.iaea.org/https://nucleus.iaea.org/sites/nss-oui/Figures/i_8960bfd7-fd63-4f8f-a6b1-ee89e5ede97b.JPG" style="width: 480px;">

Definition

The taking apart, disassembling and tearing down of the structures, systems and components of a facility for the purposes of decommissioning.

Scope notes

The two main types of dismantling are immediate dismantling and deferred dismantling.

Definition

The complete or partial removal of contamination by a deliberate physical, chemical or biological process.

Scope notes

This definition is intended to include a wide range of processes for removing contamination from people, equipment and buildings, but to exclude the removal of radionuclides from within the human body or the removal of radionuclides by natural weathering or migration processes, which are not considered to be decontamination. See also remediation.

Definition

1. An activity where types of waste or material (radioactive or exempt) are separated or are kept separate on the basis of radiological, chemical and/or physical properties, to facilitate waste handling and/or processing.

Definition

1. Determination of the nature and activity of radionuclides present in a specified place.

Scope notes

For example, characterization is the determination of the radionuclides present in a bioassay sample or in an area contaminated with radioactive material (e.g. as a first step in planning remediation). For the latter example, care should be taken to avoid confusion with the existing, and different, definition of the term site characterization.

Definition

Any or all of the operations prior to wastetreatment, such as collection, segregation, chemical adjustment and decontamination.

Definition

Operations intended to benefit safety and/or economy by changing the characteristics of the waste. Three basic treatment objectives are:
(a) Volume reduction;
(b) Removal of radionuclides from the waste;
(c) Change of composition.
Treatment may result in an appropriate waste form.

Scope notes

If treatment does not result in an appropriate waste form, the waste may be immobilized.

Definition

Those operations that produce a waste package suitable for handling, transport, storage and/or disposal.

Scope notes

Conditioning may include the conversion of the waste to a solid waste form, enclosure of the waste in containers and, if necessary, provision of an overpack.

Definition

The holding of radioactive sources, radioactive material, spent fuel or radioactive waste in a facility that provides for their/its containment, with the intention of retrieval.

Scope notes

a. Generalized from the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management [Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, INFCIRC/546, IAEA, Vienna (1997).], the Code of Conduct on the Safety and Security of Radioactive Sources [Code of Conduct on the Safety and Security of Radioactive Sources, IAEA, Vienna (2004).] and [INTERNATIONAL ATOMIC ENERGY AGENCY, Predisposal Management of Radioactive Waste, IAEA Safety Standards Series No. GSR Part 5, IAEA, Vienna (2009).].

b. Caution: Storage is by definition an interim measure, and the term interim storage would therefore be appropriate only to refer to short term temporary storage when contrasting this with the longer term fate of the waste.

c. Caution: Storage as defined above should not be described as interim storage.

d. Caution: In many cases, the only element of this definition that is important is the distinction between disposal (with no intent to retrieve) and storage (with intent to retrieve).

f. For storage in a combined storage and disposal facility, for which a decision may be made at the time of its closure whether to remove the waste stored during the operation of the storage facility or to dispose of it by encasing it in concrete, the question of intention of retrieval may be left open until the time of closure of the facility.

g. Contrasted with disposal.

e. In such cases, a definition is not necessary; the distinction can be made in the form of a footnote at the first use of the term disposal or storage (e.g. “Use of the term disposal indicates that there is no intention to retrieve the waste. If retrieval of the waste at any time in the future is intended, the term storage is used.”).

Definition

Postulated accident conditions that are not considered for design basis accidents, but that are considered in the design process of the facility in accordance with best estimate methodology, and for which releases of radioactive material are kept within acceptable limits.

Scope notes

For nuclear power plants and research reactors, design extension conditions comprise conditions in events without significant fuel degradation and conditions in events with melting of the reactor core.

Definition

Analysis using, for key parameters, single numerical values (taken to have a probability of 1), leading to a single value for the result.

Scope notes

b. Typically used with either ‘best estimate’ or ‘conservative’ values, based on expert judgement and knowledge of the phenomena being modelled.

c. Contrasting terms: probabilistic analysis or stochastic analysis. See also probabilistic analysis.

a. In the safety of nuclear installations, for example, this implies focusing on accident types, to substantiate compliance with established acceptance criteria in relation to the release of radioactive material and consequences, without considering the probabilities of different event sequences.

Definition

Probabilistic analysis is often taken to be synonymous with stochastic analysis. Strictly, however, "stochastic" conveys directly the idea of randomness (or at least apparent randomness), whereas "probabilistic" is directly related to probabilities, and hence only indirectly concerned with randomness.

Scope notes

A natural event or process might more correctly be described as "stochastic" (as in stochastic effect), whereas "probabilistic" would be more appropriate for describing a mathematical analysis of stochastic events or processes and their consequences (such an analysis would, strictly, only be "stochastic" if the analytical method itself included an element of randomness, e.g. Monte Carlo analysis).

Definition

Often used interchangeably with assessment, especially in more specific terms such as ‘safety analysis’. In general, however, analysis suggests the process and result of a study aimed at understanding the subject of the analysis, while assessment may also include determinations or judgements of acceptability. Analysis is also often associated with the use of a specific technique. Hence, one or more forms of analysis may be used in assessment.

Definition

A single action taken by a safety actuation system.

Scope notes

For example, insertion of a control rod, closing of containment valves or operation of the safety injection pumps.

Definition

Failures of two or more structures, systems or components due to a single specific event or cause.

Scope notes

a. For example, the single specific event or cause of failures (which may be failures of different types) could be a design deficiency, a manufacturing deficiency, operation and maintenance errors, a natural phenomenon, a human induced event, saturation of signals, or an unintended cascading effect from any other operation or failure within the plant or from a change in ambient conditions.

b. Common causes may be internal or external to a system.

Definition

Events unconnected with the operation of a facility or the conduct of an activity that could have an effect on the safety of the facility or activity.

Scope notes

a. Typical examples of external events for nuclear facilities include earthquakes, tornadoes, tsunamis and aircraft crashes.

b. In the case of safety assessment for long term safety in waste management, a relevant external event is one that could have an effect on the functioning of multiple barriers.

Definition

Analysis of those factors at a site that could affect the safety of a facility or activity on that site.

Scope notes

a. This includes site characterization, consideration of factors that could affect safety features of the facility or activity so as to result in a release of radioactive material and/or could affect the dispersion of such material in the environment, as well as population and access issues relevant to safety (e.g. feasibility of evacuation, location of people and resources).

b. The analysis for a site of the origins of external events that could give rise to hazards with potential consequences for the safety of a nuclear power plant constructed on that site.

c. For a nuclear power plant, site evaluation typically involves the following stages:
(a) Site selection stage. One or more preferred candidate sites are selected after the investigation of a large region, the rejection of unsuitable sites, and screening and comparison of the remaining sites.
(b) Site characterization stage. This stage is further subdivided into: - Site verification, in which the suitability of the site to host a nuclear power plant is verified mainly according to predefined site exclusion criteria; - Site confirmation, in which the characteristics of the site necessary for the purposes of analysis and detailed design are determined.
(c) Pre-operational stage. Studies and investigations begun in the previous stages are continued after the start of construction and before the start of operation of the plant, to complete and refine the assessment of site characteristics. The site data obtained allow a final assessment of the simulation models used in the final design.
(d) Operational stage. Appropriate safety related site evaluation activities are carried out throughout the lifetime of the facility, mainly by means of monitoring and periodic safety review.

Definition

An analytical or physical representation or quantification of a real system and the ways in which phenomena occur within that system, used to predict or assess the behaviour of the real system under specified (often hypothetical) conditions.

Definition

1. Methods or physical structures designed to prevent or control the release and the dispersion of radioactive substances.

Scope notes

Although related to confinement, containment is usually used to refer to methods or structures that perform a confinement function in facilities and activities, namely preventing or controlling the release of radioactive substances and their dispersion in the environment. See confinement for a more extensive discussion.

In the context of waste disposal, the containment of the radionuclides associated with the waste is through the provision of engineered barriers and natural barriers - including the waste form and packaging, backfill materials, the host environment and geological formations - for confinement of the radionuclides within the waste matrix, the packaging and the disposal facility and thus its isolation from the environment.

Definition

Plant state, following an anticipated operational occurrence or accident conditions, in which fulfilment of the fundamental safety functions can be ensured and which can be maintained for a time sufficient to implement provisions to reach a safe state.

Definition

Separation by geometry (distance, orientation, etc.), by appropriate barriers, or by a combination thereof.

Definition

The physical separation and retention of radioactive waste away from people and from the environment.

Scope notes

Isolation of radioactive waste with its associated hazards in a disposal facility involves the minimization of the influence of factors that could reduce the integrity of the disposal facility; provision for a very low mobility of most long lived radionuclides to impede their migration from the disposal facility; and making access to the waste by people difficult without special technical capabilities.

Design features are intended to provide isolation (a confinement function) for several hundreds of years for short lived waste and for at least several thousand years for intermediate level waste and high level waste. Isolation is an inherent feature of geological disposal.

Definition

For nuclear power plants, the classes into which systems and components and other items of equipment are assigned on the basis of their functions and their safety significance.

Scope notes

a. The design is required to ensure in particular that any failure of items important to safety in a system in a lower safety class will not propagate to a system in a higher safety class. Items of equipment that perform multiple functions are required to be classified in a safety class that is consistent with the most important function performed by the items of equipment.

Definition

The manner or state in which a structure, system or component fails.

Definition

An action on a non-conforming product to make it acceptable for its intended use (see ISO 9000:2015, INTERNATIONAL ORGANIZATION FOR STANDARDIZATION, Quality Management Systems — Fundamentals and Vocabulary, ISO 9000:2015, ISO, Geneva (2015).).

Definition

The presence of two or more independent (redundant) systems or components to perform an identified function, where the different systems or components have different attributes so as to reduce the possibility of common cause failure, including common mode failure.

Scope notes

Examples of such attributes are: different operating conditions different working principles or different design teams (which provide functional diversity), and different sizes of equipment, different manufacturers, and types of equipment (which provide diversity of equipment) that use different physical methods (which provide physical diversity).

When the term diversity is used with an additional attribute, the term diversity indicates the general meaning ‘existence of two or more different ways or means of achieving a specified objective’, while the attribute indicates the characteristics of the different ways applied, e.g. functional diversity, equipment diversity, signal diversity, functional diversity. Application of diversity at the level of functions in applications in process engineering (e.g. for the actuation of a trip on both a pressure limit and a temperature limit).

Definition

Provision of alternative (identical or diverse) structures, systems and components, so that any single structure, system or component can perform the required function regardless of the state of operation or failure of any other.

Definition

A criterion (or requirement) applied to a system such that it must be capable of performing its task in the presence of any single failure.

Scope notes

To ensure that the single failure criterion is met, usually two or more independent (redundant) systems or trains are provided by design to achieve the same safety function.

Definition

The assembly of equipment designated to perform all actions required for a particular initiating event to ensure that the limits specified in the design basis for anticipated operational occurrences and design basis accidents are not exceeded.

Scope notes

Caution: The term ‘group’ is also used (with various qualifying adjectives, e.g. maintenance group, commissioning group) in the more obvious sense of a group of people involved in a particular area of work. Such terms may need to be defined if there is any chance of confusion with safety group.

Definition

A failure which results in the loss of capability of a single system or component to perform its intended safety function(s), and any consequential failure(s) which result from it.

Definition

A component whose functioning does not depend on an external input such as actuation, mechanical movement or supply of power.

Scope notes

See also component, core components and structures, systems and components.

a. A passive component has no moving part, and, for example, only experiences a change in pressure, in temperature or in fluid flow in performing its functions. In addition, certain components that function with very high reliability based on irreversible action or change may be assigned to this category.

b. Examples of passive components are heat exchangers, pipes, vessels, electrical cables and structures. It is emphasized that this definition is necessarily general in nature, as is the corresponding definition of active component.

c. Certain components, such as rupture discs, check valves, safety valves, injectors and some solid state electronic devices, have characteristics which require special consideration before designation as an active component or a passive component.

d. Any component that is not a passive component is an active component.

Definition

Limits on operational parameters within which an authorized facility has been shown to be safe.

Scope notes

Safety limits are beyond the limits for normal operation.

Definition

1. A type of inspection to verify the integrity of a facility or structure.

Scope notes

For example, surveillance is used in the context of a disposal facility for radioactive waste to mean physical inspection of the facility to verify its integrity and the capability to protect and preserve passive barriers.

Definition

1. An examination, observation, surveillance, measurement or test undertaken to assess structures, systems and components and materials, as well as operational activities, technical processes, organizational processes, procedures and personnel competence.

Definition

A system important to safety that is not part of a safety system or a safety feature for design extension conditions.

Scope notes

The reactor coolant system, for example, is an item important to safety that is neither a safety system nor a safety feature for design extension conditions.

Definition

A set of operations that establish, under specified conditions, the relationship between values of quantities indicated by a measuring instrument or measuring system, or values represented by a material measure or a reference material, and the corresponding values realized by standards.

Scope notes

b. A calibration may be expressed by a statement, calibration function, calibration diagram, calibration curve or calibration table. In some cases, it may consist of an additive or multiplicative correction of the indication with associated measurement uncertainty.

c. Calibration should not be confused with adjustment of a measuring system, often mistakenly called ‘self-calibration’, or with verification of calibration.

a. The definition was previously “A measurement of, or adjustment to, an instrument, component or system to ensure that its accuracy or response is acceptable”.

Definition

Periodic testing to verify that structures, systems and components continue to function or are capable of performing their functions when called upon to do so.

Definition

Process of determining whether a system or component is suitable for operational use.

Scope notes

a. Qualification is generally performed in the context of a specific set of qualification requirements for the specific facility and class of system and for the specific application.

c. Qualification may rely on activities performed outside the framework of a specific facility design (this is called ‘generic qualification’ or ‘prequalification’).

d. Prequalification may significantly reduce the necessary effort in facility specific qualification; however, the application specific qualification requirements must still be met and be shown to be met.

b. Qualification may be accomplished in stages: for example, first, by the qualification of pre-existing equipment (usually early in the system realization process), then, in a second step, by the qualification of the integrated system (i.e. in the final realized design).

e. See also harsh environment, mild environment.

Definition

General process in which characteristics of a structure, system or component gradually change with time or use.

Scope notes

Although the term ageing is defined in a neutral sense — the changes involved in ageing may have no effect on protection or safety, or could even have a beneficial effect — it is most commonly used with a connotation of changes that are (or could be) detrimental to protection and safety (i.e. as a synonym of ageing degradation).

Definition

The period from initial operation to final withdrawal from service of a structure, system or component.

Definition

Individual workers engaged in the operation of an authorized facility or the conduct of an authorized activity.

Scope notes

Caution: This may be shortened to operator(s), provided that there is no danger of confusion with operator in the sense of operating organization.

Definition

1. Material containing any of the fissile nuclides in sufficient proportion to enable a self-sustained nuclear chain reaction with slow (thermal) neutrons.

Definition

The state of a nuclear chain reacting medium when the chain reaction is just self-sustaining (or critical), i.e. when the reactivity is zero.

Scope notes

Often used, slightly more loosely, to refer to states in which the reactivity is greater than zero.

Definition

A non-routine situation or eventthat necessitates prompt action, primarily to mitigate a hazard or adverse consequences for human life, health, property and the environment.

Scope notes

a. This includes nuclear and radiological emergencies and conventional emergencies such as fires, release of hazardous chemicals, storms or earthquakes.

b. It includes situations for which prompt action is warranted to mitigate the effects of a perceived hazard.

c. All terms and definitions relating to an emergency are taken from [FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS, INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL CIVIL AVIATION ORGANIZATION, INTERNATIONAL LABOUR ORGANIZATION, INTERNATIONAL MARITIME ORGANIZATION, INTERPOL, OECD NUCLEARENERGY AGENCY, PAN AMERICAN HEALTH ORGANIZATION, PREPARATORY COMMISSION FOR THE COMPREHENSIVE NUCLEAR-TESTBAN TREATY ORGANIZATION, UNITED NATIONS ENVIRONMENT PROGRAMME, UNITED NATIONS OFFICE FOR THE COORDINATION OF HUMANITARIAN AFFAIRS, WORLD HEALTH ORGANIZATION, WORLD METEOROLOGICAL ORGANIZATION, Preparedness and Response for a Nuclear or Radiological Emergency, IAEA Safety Standards Series No. GSR Part 7, IAEA, Vienna (2015).].

Definition

aleatory uncertainty. Uncertainty inherent in a phenomenon.
epistemic uncertainty. Uncertainty attributable to incomplete knowledge about a phenomenon, which affects the ability to model it.

Definition

3. Determination of the character of something.

Scope notes

This is the standard dictionary definition and would not need to be included in an individual glossary. It is included here only to distinguish the usual usage from the more restricted usage indicated in (1) and (2).

Definition

A rod of nuclear fuel, its cladding and any associated components necessary to form a structural entity.

Scope notes

Commonly referred to as a fuel rod in light water reactors.

Definition

A radionuclide produced by nuclear fission.

Scope notes

Used in contexts where the radiation emitted by the radionuclide is the potential hazard.

Definition

A set of fuel elements and associated components which are loaded into and subsequently removed from a reactor core as a single unit.

Definition

A planar or gently curved fracture surface or zone of the Earth across which there has been relative displacement.

Definition

1. The quantity A for an amount of radionuclide in a given energy state at a given time, defined as
A(t) = dN / dt
where dN is the expectation value of the number of spontaneous nuclear transformations from the given energy state in the time interval dt.

Scope notes

b. The SI unit for activity is reciprocal second (s^–1), termed the becquerel (Bq).

c. Formerly expressed in [curie] (Ci); activity values may be given in Ci (with the equivalent in Bq in parentheses) if they are being quoted from a reference that uses Ci as the unit.

a. The rate at which nuclear transformations occur in a radioactive material. The equation is sometimes given as:
A(t) = dN / dt
where N is the number of nuclei of the radionuclide, and hence the rate of change of N with time is negative. Numerically, the two forms are identical.

Definition

The sum of the heat originating from radioactive decay and shutdown fission and the heat stored in reactor related structures and in heat transport media.

Definition

2. Typically, the tube of material that houses nuclear fuel pellets and provides the containment (means of confinement) of radionuclides produced during fission.

Scope notes

Cladding may also provide structural support.

The cladding tube, together with the end cups or plugs, also typically provides structural support.

Definition

1. A structurally closed physical barrier (especially in a nuclear installation) designed to prevent or control the release and the dispersion of radioactive substances, and its associated systems.

Definition

System that monitors the operation of a reactor and which, on sensing an abnormal condition, automatically initiates actions to prevent an unsafe or potentially unsafe condition.

Scope notes

Caution: This use of the term protection refers to protection of the plant (protection (2)).

The system in this case encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device input terminals.

Definition

Application of diversity at the level of functions in applications in process engineering (e.g. for the actuation of a trip on both a pressure limit and a temperature limit).

Definition

An arrangement of interconnected components within a system that initiates a single output.

Scope notes

A channel loses its identity where single output signals are combined with signals from other channels (e.g. from a monitoring channel or a safety actuation channel).

The above definition is specific to a particular area of nuclear safety. The term channel is also used in its usual senses (and therefore usually without specific definition) in a variety of contexts.

Definition

An organization designated or recognized by a State as being responsible for managing or implementing any aspect of an emergency response.

Scope notes

This also includes those organizations or services necessary to support the management and/or conduct of an emergency response, such as meteorological services.

Definition

1. Nuclear fuel removed from a reactor following irradiation that is no longer usable in its present form because of depletion of fissile material, poison buildup or radiation damage.

Scope notes

The participle ‘spent’ suggests that spent fuel cannot be used as fuel in its present form (e.g. as in spent source). In practice, however (as in (2) below), spent fuel is commonly used to refer to fuel that has been used as fuel but will no longer be used, whether or not it could be used (and that might more accurately be termed ‘disused fuel’).

Definition

A component that converts energy into action when commanded by an actuation device.

Scope notes

Such as a motor, solenoid operator or pneumatic operator.

Definition

A component that directly controls the motive power for actuated equipment.

Scope notes

Examples of actuation devices include circuit breakers and relays that control the distribution and use of electric power, and pilot valves controlling hydraulic or pneumatic fluids.

Definition

1. The measurement of dose, dose rate or activity for reasons relating to the assessment or control of exposure to radiation or exposure due to radioactive substances, and the interpretation of the results.

Scope notes

a. ‘Measurement’ is used somewhat loosely. The ‘measurement’ of dose often means the measurement of a dose equivalent quantity as a proxy (i.e. substitute) for a dose quantity that cannot be measured directly. Also, sampling may be involved as a preliminary step to measurement.

d. Monitoring may be subdivided in two different ways: according to where the measurements are made, into individual monitoring, workplace monitoring, source monitoring and environmental monitoring; and, according to the purpose of the monitoring, into routine monitoring, task related monitoring and special monitoring.

b. Measurements may actually be of radiation levels, airborne activity concentrations, levels of contamination, quantities of radioactive material or individual doses.

c. The results of these measurements may be used to assess radiological hazards or doses resulting or potentially resulting from exposure.

Definition

The spreading of radionuclides in air (aerodynamic dispersion) or water (hydrodynamic dispersion) resulting mainly from physical processes affecting the velocity of different molecules in the medium.

Scope notes

a. Often used in a more general sense combining all processes (including molecular diffusion) that result in the spreading of a plume. The terms atmospheric dispersion and hydrodynamic dispersion are used in this more general sense for plumes in air and water, respectively.

b. In usual language synonymous with dispersal, but dispersion is mostly used more specifically as defined above, whereas dispersal is typically (though not universally) used as a more general expression.

Definition

This terms encompasses several sub-terms, such as [radiation area], controlled area, hazard control area, inner area, limited access area, operational control area, operations area, protected area, site area, site boundary, supervised area.

Definition

1. Emplacement of waste in an appropriate facility without the intention of retrieval.

Scope notes

c. Caution: In IAEA publications, disposal should be used only in accordance with the more restrictive definition given above.

d. Caution: In many cases, the only element of this definition that is important is the distinction between disposal (with no intent to retrieve) and storage (with intent to retrieve). In such cases, a definition is not necessary; the distinction can be made in the form of a footnote at the first use of the term disposal or storage (e.g. “The use of the term ‘disposal’ indicates that there is no intention to retrieve the waste. If retrieval of the waste at any time in the future is intended, the term ‘storage’ is used.”).

e. Caution: The term disposal implies that retrieval is not intended and would require deliberate action to regain access to the waste; it does not mean that retrieval is not possible.

g. Contrasted with storage.

f. For storage in a combined storage and disposal facility, for which a decision may be made at the time of its closure whether to remove the waste stored during the operation of the storagefacility or to dispose of it by encasing it in concrete, the question of intention of retrieval may be left open until the time of closure of the facility.

b. In some States, the term disposal is used administratively in such a way as to include, for example, incineration of waste or the transfer of waste between operators.

a. In some States, the term disposal is used to include discharges of effluents to the environment.

Definition

New fuel or unirradiated fuel, including fuel fabricated from fissionable material recovered by reprocessing previously irradiated fuel.

Definition

A prospective and source related value of individual dose that is used in planned exposure situations as a parameter for the optimization of protection and safety for the source, and that serves as a boundary in defining the range of options in optimization.

Scope notes

c. The dose constraint for each particular source is intended, among other things, to ensure that the sum of doses from planned operations for all sources under control remains within the dose limit.

a. For occupational exposure, the dose constraint is a constraint on individual dose to workers established and used by registrants and licensees to set the range of options in optimizing protection and safety for the source.

b. For public exposure, the dose constraint is a source related value established or approved by the government or the regulatory body, with account taken of the doses from planned operations of all sources under control.

d. For medical exposure, the dose constraint is a source related value used in optimizing the protection of carers and comforters of patients undergoing radiological procedures, and the protection of volunteers subject to exposure as part of a programme of biomedical research.

Definition

[A radiation generator, or a radioactive source or other radioactive material outside the nuclear fuel cycles of research and power reactors.]

Scope notes

Caution: Defined in the 2001 edition of the Code of Conduct on the Safety and Security of Radioactive Sources, but not included in the 2004 edition (see Code of Conduct on the Safety and Security of Radioactive Sources, IAEA, Vienna (2004).).

Definition

2. The movement of something as a result of being carried by a medium.

Scope notes

A general term used when a number of different processes are involved. The most common examples are heat transport — a combination of advection, convection, etc., in a cooling medium — and radionuclide transport in the environment — which could include processes such as advection, diffusion, sorption and uptake.

Definition

A radionuclide produced by activation.

Scope notes

Often used to distinguish from fission products. For example, in decommissioningwaste comprising structural materials from a nuclear facility, activation products might typically be found primarily within the matrix of the material, whereas fission products are more likely to be present in the form of contamination on surfaces.

<h2>Definition</h2><img src="https://vocabulary.iaea.org/https://nucleus.iaea.org/sites/nss-oui/Figures/i_2d62d442-372e-46ef-9384-c7a0f98dcf19.JPG" style="width: 800px;"> 

Definition

1. The dose per unit time.

Scope notes

Caution: Although dose rate could, in principle, be defined over any unit of time (e.g. an annual dose is technically a dose rate), in IAEA publications the term dose rate should be used only in the context of short periods of time, for example dose per second or dose per hour.

Definition

A defined area in which specific protection measures and safety provisions are or could be required for controlling exposures or preventing the spread of contamination in normal working conditions, and preventing or limiting the extent of potential exposures.

Scope notes

A controlled area is often within a supervised area, but need not be.

The term [radiation area] is sometimes used to describe a similar concept, but controlled area is preferred in IAEA publications.

Definition

The measurement of external dose rates due to sources in the environment or of radionuclide concentrations in environmental media.

Scope notes

Contrasted with source monitoring.

Definition

A route by which radiation or radionuclides can reach humans and cause exposure.

Scope notes

An exposure pathway may be very simple, for example the external exposurepathway from airborne radionuclides, or a more complex chain, for example the internal exposurepathway from drinking milk from cows that ate grass contaminated with deposited radionuclides.

Definition

Any substance, whether processed, semi-processed or raw, that is intended for human consumption.

Scope notes

This includes foodstuffs and drink (other than fresh water), chewing gum and substances used in the preparation or processing of food; it does not include cosmetics, tobacco or drugs. Consumption in this context refers to ingestion.

Definition

Settings for levels at which safety systems are automatically actuated in the event of anticipated operational occurrences or design basis accidents, to prevent safety limits from being exceeded.

Operational states		Accident conditions
Normal operation	Anticipated operational occurences	Design basis accidents	Design extension conditions
			Without significant fuel degradation	With core melting