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Safety of Research Reactors
Safety of Research Reactors
1INTERNATIONAL ATOMIC ENERGY AGENCY, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005).
2The important areas of research reactor safety include all activities performed to achieve the purpose for which the research reactor was designed and constructed or modified. This includes: maintenance, testing and inspection; fuel handling and handling of radioactive material (including the production of radioisotopes); the installation, testing and operation of experimental devices; the use of neutron beams; research and development work and education and training using the research reactor systems; and other associated activities.
3A research reactor is a nuclear reactor used mainly for the generation and utilization of neutron flux and ionizing radiation for research and other purposes, including experimental facilities associated with the reactor and storage, handling and treatment facilities for radioactive materials on the same site that are directly related to safe operation of the research reactor. Facilities commonly known as critical assemblies and subcritical assemblies are included.
4Within this context, the site area is the geographical area that contains an authorized facility, authorized activity or radiation source, and within which the management of the authorized facility or authorized activity may directly initiate emergency actions. The site boundary is the perimeter of the site area.
5For the purposes of this safety standard, the term experimental devices includes devices installed in or around a reactor to utilize the neutron flux and ionizing radiation from the reactor for research, development, isotope production or any other purpose.
7Authorization to operate a facility or to conduct an activity may be granted by the regulatory body or by another governmental body to an operating organization or to a person. ‘Authorization’ includes approval, written permission, licensing, certification or registration. See Ref.  and Requirement 23 of GSR Part 1 (Rev. 1) .
9Although the utilization and modification of research reactors are activities that are normally included under operation, they may be considered separate stages in the authorization process, since their safety implications give rise to a large number of review and assessment activities that are repeated many times over the lifetime of the reactor facility (see paras 7.98–7.106).
10The operating personnel comprise the reactor manager, the shift supervisors, the operators, the maintenance staff and the radiation protection staff.
11Independent assessments such as audits or surveillance are carried out to determine the extent to which the requirements for the management system are fulfilled, to evaluate the effectiveness of the management system and to identify opportunities for improvement. They can be conducted by or on behalf of the organization itself for internal purposes, by interested parties such as customers and regulators (or by other persons on their behalf), or by independent external organizations.
12‘Senior management’ means the person who, or group of people that, is accountable for meeting the terms established in the licence and directs, controls and assesses an organization at the highest level. Many different terms are used, including, for example: board of directors, chief executive officer, director general, executive team, plant manager, top manager, chief regulator, site vice-president, managing director and laboratory director.
13An integrated management system is a single coherent management system in which all constituents of an organization are integrated to enable the organization’s objectives to be achieved. Such constituents include the organizational structure, resources and organizational processes. This system integrates all elements of management, including safety, health, environmental, security, quality, human–and-organizational-factor, societal and economic elements, so that safety is not compromised.
14Resources includes individuals, infrastructure, the working environment, information and knowledge, and suppliers, as well as material and financial resources.
15Requirements for safety assessment for facilities and activities are established in IAEA Safety Standards Series No. GSR Part 4 (Rev. 1), Safety Assessment for Facilities and Activities .
16A peer review is a review conducted by a team of independent experts with technical competence and experience in the areas of evaluation. Judgements are based on the combined expertise of the team members. The objectives, scope and size of the review team are tailored to the review that is to be conducted. A review is neither an inspection nor an audit against specific standards. Instead, it consists of a comprehensive comparison of the practices applied by organizations with internationally accepted good practices, and an exchange of expert judgement.
17In some States, an additional safety committee (or advisory group) is established to advise the reactor manager on the safety aspects of the day to day operation and utilization of the reactor (see para. 7.26).
18The nuclear fuel elements are the elements containing fissionable and fissile nuclear material that are used in the core of a research reactor for the purpose of generating neutrons. Adequate design and safety margins are established to take into account unknown behaviour of experimental fuel that is not yet qualified.
19The reactor management comprises the members of the operating organization to whom the responsibility and the authority for directing the operation of the research reactor facility have been assigned.
20The possibility of certain conditions occurring is considered to have been practically eliminated (i.e. eliminated from further consideration) if it is physically impossible for the conditions to occur or if the conditions can be considered with a high level of confidence to be extremely unlikely to arise.
21Requirements on radiation protection and the safety of radiation sources for facilities and activities are established in GSR Part 3 .
22Safety classification reflects the significance for nuclear safety of the structures, systems and components. Its purpose is to establish a grading in the application of the requirements for design. There are other possible classifications or categorizations of structures, systems and components in accordance with other aspects (e.g. seismic or environmental qualification, or quality categorization of structures, systems and components).
23This aspect is important in particular for critical assemblies and subcritical assemblies and dry fuel storage facilities, which shall be designed to be safely subcritical following activation of the fire protection system and during firefighting activities.
24The analysis of design extension conditions could be performed by means of a best estimate approach (more stringent approaches may be used according to States’ requirements).
25Confinement is the prevention or control of releases of radioactive material to the environment in operation or in accidents . Confinement is a basic safety function that is required to be fulfilled in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected design extension conditions. The function of confinement is usually fulfilled by means of several barriers surrounding the main parts of a nuclear reactor that contain radioactive material. For a research reactor, the reactor building may be the ultimate barrier for ensuring confinement. Consideration may be given to the use of other structures (e.g. the reactor block in a fully enclosed research reactor) for providing confinement where this is technically feasible. For most designs of large nuclear reactor, a strong structure housing the reactor is the ultimate barrier providing confinement. Such a structure is called the containment structure or simply the containment. The containment also protects the reactor against external events and provides radiation shielding in operational states and in accident conditions.
26An early radioactive release is a release for which off-site protective measures are necessary but are unlikely to be fully effective in due time. A large radioactive release is a release for which off-site protective measures limited in terms of times and areas of application are insufficient to protect people and the environment.
27The safety margin is the difference between the safety limit and the operational limit. It is sometimes expressed as the ratio of these two values.
28Including means of communication within the supplementary control room, if one exists.
29Reactivity control mechanisms are devices of all kinds for controlling the reactivity, including regulating rods, control rods, shutdown rods or blades, and devices for controlling the moderator level or the reflection.
30The cooling requirement might not apply to some types of critical assembly and subcritical assembly.
31A subcritical assembly can be ‘shut down’ by the withdrawal of the neutron source.
32The shutdown margin is the negative reactivity provided in addition to the negative reactivity necessary to maintain the reactor in a subcritical condition without time limit, with the most reactive control device removed from the core and with all experiments that can be moved or changed during operation in their most reactive condition.
34A flapper is a passive valve that opens when the flow (pressure) is below a set value to allow for the creation of natural circulation in the event of a loss of forced flow.
36Emergency response facilities and locations are addressed in GSR Part 7 . For research reactors, emergency response facilities (which are separate from the control room and the supplementary control room) include the emergency centre, and the technical support centre and the operational support centre, as appropriate.
37Operation includes all activities performed to achieve the purpose for which the nuclear research reactor was designed and constructed or modified. Besides operating the reactor, this includes: maintenance, testing and inspection; fuel handling and handling of radioactive material, including the production of radioisotopes; installation, testing and operation of experimental devices; the use of neutron beams; the use of the research reactor systems for the purposes of research and development and education and training; and other associated activities.
38The reactor manager is the member of the reactor management to whom the direct responsibility and authority for the safe operation of the research reactor is assigned by the operating organization and whose primary duties comprise the fulfilment of this responsibility.
39The reactor manager does not necessarily need to hold a licence to operate the reactor, but needs to have completed a training programme (see para. 7.30).
40Facilities of low potential hazard might not need to have these positions. However, the functions need to be covered within such facilities.
41Initial criticality tests and low power tests and Stage C of the commissioning programme might not apply to subcritical assemblies, providing adequate subcriticality has been verified (e.g. through 1/M calculations, where M is the subcritical neutron multiplication factor).
42Normal operation is operation within specified operational limits and conditions. For a research reactor, this includes startup, low and nominal power operation, shutting down, shutdown, maintenance, testing and refuelling.
43Emergency procedures are developed as an element of separate emergency arrangements (see paras 7.89–7.93) and in accordance with GSR Part 7 .
44Low power research reactors and subcritical assemblies usually have a lifetime core, which could be specified in the operational limits and conditions in terms of factors other than burnup (e.g. completion of the experimental programme). Nevertheless, the value of the maximum burnup is one of the parameters that is considered in the determination of the core lifetime.
45‘Non-radiation-related safety’ concerns hazards other than radiation related hazards; this is sometimes referred to as industrial safety or conventional safety.
46Part of this process for the characterization, classification, processing, transport, storage and disposal of radioactive waste could be carried out by another organization.
47Periodic safety review is 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) .
48A research reactor in extended shutdown is one that is no longer operating, with no decision on its decommissioning, and where there is no clear decision about the future of the reactor as to whether it will be brought back into operation or decommissioned. Long shutdown periods for maintenance or for implementation of refurbishment and modification projects are not considered an extended shutdown state.
50Although a loss of normal electrical power is not considered an initiating event, consideration has to be given to the loss of normal electrical power followed by the loss of emergency power to ensure that the consequences would be acceptable under emergency conditions (e.g. a drop in voltage may cause devices to fail at different times).
Tags applicable to this publication
- Publication type:Specific Safety Requirements
- Publication number: SSR-3
- Publication year: 2016