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April 2017
IAEA SAFETY STANDARDS
for protecting people and the environment
Status: Step 8 - for submission to Members
States for 120-day comment
Deadline for comments: 18 August 2017
Radiation Safety of Accelerator Radioisotope
Production Facilities
DRAFT SAFETY GUIDE No. SSGDS434
IAEA
International Atomic Energy Agency
CONTENTS
1.
INTRODUCTION ............................................................................................................ 1
BACKGROUND .............................................................................................................. 1
OBJECTIVE ..................................................................................................................... 2
SCOPE 2
STRUCTURE ................................................................................................................... 3
2.
JUSTIFICATION OF PRACTICES ................................................................................. 4
3.
TYPES OF RADIOISOTOPE PRODUCTION FACILITIES ......................................... 5
4.
DUTIES AND RESPONSIBILITIES .............................................................................. 6
GENERAL ........................................................................................................................ 6
RADIATION PROTECTION OFFICER ......................................................................... 9
QUALIFIED EXPERTS/RADIATION PROTECTION ADVISER ............................. 11
WORKERS ..................................................................................................................... 13
LOCAL RULES AND PROCEDURES ......................................................................... 14
5.
SAFETY ASSESSMENT ............................................................................................... 15
GENERAL ...................................................................................................................... 15
6.
RADIATION PROTECTION PROGRAMME.............................................................. 24
GENERAL ...................................................................................................................... 24
STRUCTURE OF THE RADIATION PROTECTION PROGRAMME ...................... 25
MANAGEMENT STRUCTURE AND POLICIES ....................................................... 25
HEALTH SURVEILLANCE PROGRAMME .............................................................. 28
RADIATION SAFETY COMMITTEE ......................................................................... 29
7.
TRAINING AND EDUCATION ................................................................................... 29
GENERAL ...................................................................................................................... 29
TRAINING PROGRAMME .......................................................................................... 30
STRUCTURE AND CONTENT OF THE TRAINING COURSE ................................ 31
REFRESHER TRAINING ............................................................................................. 33
8.
INDIVIDUAL MONITORING OF WORKERS ........................................................... 34
INDIVIDUAL DOSE ASSESSMENT AND RECORD KEEPING .............................. 34
EXTERNAL DOSIMETRY ........................................................................................... 35
INTERNAL DOSIMETRY ............................................................................................ 37
INVESTIGATION OF DOSES ...................................................................................... 38
9.
WORKPLACE MONITORING ..................................................................................... 39
10.
ENVIRONMENTAL MONITORING AND EFFLUENT DISCHARGE .................... 45
ENVIRONMENTAL MONITORING ........................................................................... 45
EFFLUENT DISCHARGE............................................................................................. 45
AIR EFFLUENT MONITORING .................................................................................. 46
LIQUID EFFLUENT MONITORING ........................................................................... 47
MINIMIZING EFFLUENT DISCHARGES .................................................................. 48
11.
PERSONAL PROTECTIVE EQUIPMENT .................................................................. 50
12.
NUCLEAR SECURITY CONSIDERATIONS ............................................................. 52
13.
TESTING AND MAINTENANCE OF EQUIPMENT.................................................. 53
PERIODIC TESTS ......................................................................................................... 54
RECORDS ...................................................................................................................... 55
FACILITY MAINTENANCE AND MODIFICATION ................................................ 55
14.
RADIOACTIVE WASTE MANAGEMENT AND DECOMMISSIONING ................ 56
CHARACTERIZATION OF RADIOACTIVE WASTE ............................................... 57
PRINCIPLES OF WASTE MINIMIZATION ............................................................... 58
HANDLING AND PROCESSING OF RADIOACTIVE WASTE ............................... 58
OTHER HANDLING GUIDELINES ............................................................................ 59
ON-SITE STORAGE OF RADIOACTIVE WASTE .................................................... 59
PREPARATION OF WASTE SHIPMENTS ................................................................. 60
15.
TRANSPORT OF RADIOACTIVE MATERIAL ......................................................... 60
TRANSPORT REQUIREMENTS ................................................................................. 60
16.
EMERGENCY PREPAREDNESS AND RESPONSE .................................................. 62
REFERENCES ......................................................................................................................... 68
ANNEX I KEY RADIATION PROTECTION ISSUES TO BE TAKEN INTO
ACCOUNT WHEN PLANNING THE PRODUCTION OF CYCLOTRON
RADIOISOTOPES ......................................................................................................... 71
ANNEX II
EXAMPLE OF IMMEDIATE ON-SITE RESPONSE ACTIONS IN
CASE OF AN EMERGENCY AT A RADIOISOTOPE PRODUCTION
FACILITY ...................................................................................................................... 72
CONTRIBUTORS TO DRAFTING AND REVIEW ............................................................. 74
1. INTRODUCTION
BACKGROUND
1.1.Radionuclides are used worldwide in a range of medical, industrial, research and
academic applications that bring many benefits to humankind. Most of these radionuclides are
produced in reactors and particle accelerators. The facilities which produce radionuclides and
the facilities in which radionuclides are processed are referred to collectively as ‘radioisotope
production facilities’.1. The operation of reactors and particle accelerators and the subsequent
processing of radioactive material can present significant radiation hazards to workers,
members of the public, and the environment unless they are properly controlled.
1.2. In 2003, there were 278 research reactors in operation, of which approximately 70 were
deemed useful for regular radioisotope production [1]. In 2006, it was estimated that there
were approximately 350 cyclotrons in operation worldwide that were used to some extent for
radioisotope production [2]. The number of institutions that operate cyclotrons and
manufacture and distribute radiopharmaceuticals that are used in positron emission
tomography and single photon emission computed tomography is significant and growing.
1.3. The IAEA Safety Standards Series No. GSR Part 3, Radiation Protection and Safety of
Radiation Sources: International Basic Safety Standards [3] establish the basic requirements
for protection of people against exposure to ionizing radiation and for the safety of radiation
sources2. The implementation of these requirements at radioisotope production facilities is
intended to prevent accidents and, generally, to provide for the best possible protection and
safety measures under the prevailing circumstances. The magnitudes and likelihood of
exposures and the number of individuals exposed are required to be kept as low as reasonably
achievable, economic and social factors being taken into account.
1.4. Unless otherwise stated, terms are used with the meanings ascribed to them in the IAEA
Safety Glossary (2007 Edition) [4].
1
The word ‘radioisotope’ is commonly used in the context of the facilities considered in this Safety Guide and is
therefore retained here. Strictly, the word ‘radionuclide’ should be used or the word qualified by the name of the
element to which it relates, for example, radioisotope of cobalt.
2
The term ‘radiation source’ includes radioactive sources and radiation generators. ‘Radiation’ as used in the
IAEA Safety Standards means ionizing radiation.
1
OBJECTIVE
1.5. The objective of this Safety Guide is to provide recommendations on how to meet the
requirements of GSR Part 3 [3] with regard to radioisotope production facilities. This Safety
Guide provides specific, practical recommendations on the safe design and operation of these
facilities for use by operating organizations and the designers of these facilities, and by
regulatory bodies.
SCOPE
1.6. This Safety Guide addresses the radiation safety and protection aspects of the process
whereby radioisotopes that have been produced in accelerators (principally cyclotrons), or
purified from other sources are processed into radioactive products for subsequent use, for
example, in nuclear medicine. It also addresses elements of the design and operation of
accelerators (principally cyclotrons) that pertain directly to the production of radioisotopes.
1.7. The following types of facilities that produce radioisotopes are within the scope of this
Safety Guide:
(a)
Facilities that process targets that have been irradiated by a charged particle beam of an
accelerator to produce radioisotopes;
(b)
Accelerator facilities with energies of less than 70 MeV/nucleon that are operated
principally to produce radioisotopes. This document addresses these in the following
four categories of accelerators:
(i) Low energy (<20 MeV /nucleon) cyclotrons for medical radioisotope
production;
(ii) 20 – 40 MeV/ nucleon isotope production cyclotrons;
(iii)> 40 MeV/ nucleon cyclotrons for mixed research and radioisotope production;
(iv) Linear accelerators used for radioisotope production.
1.8. The use of radioactive material following its manufacture, and standards and quality
assurance procedures that pertain to its production are outside the scope of this document. The
production of fissile material is outside the scope of this document.
1.9.The design and operation of reactors is outside the scope of this document to avoid
duplication with a number of IAEA publications on research reactors.
2
1.10. Centralized radiopharmacies that formulate radiopharmaceuticals from bulk quantities
of radioisotopes or generators are outside the scope of this document.
1.11. Radiation generators (e.g. linear accelerators used in radiotherapy applications) that
produce radioisotopes as a by-product of their operation are outside the scope of this
document.
1.12. Consideration of non-radiological related risks and of the benefits of radioisotopes that
are produced in radioisotope production facilities are outside the scope of this Safety Guide.
1.13. The Safety Guide also provides information on the need for appropriate nuclear security
measures and on their interface with safety measures, but does not provide specific guidance
on such nuclear security aspects. Additional security guidance can be found in the IAEA’s
Nuclear Security Series [5, 6, 7, 8].
STRUCTURE
1.14. The justification of radioisotope production facilities is discussed in Section 2. Designs
of irradiation facilities are categorized according to radiation type and methods of
accessibility and shielding, as described in Section 3 of this Safety Guide. The authorization
of irradiation practices, the responsibilities of the operating organization and general radiation
safety issues are discussed in Section 4. The safety assessment duties and radiation protection
programme are described in Sections 5 and 6 respectively.
1.15. Section 7 provides a description of training and education of personnel of radioisotope
production facilities. Section 8 deals with individual monitoring of workers of isotope
production facilities. Section 9 discusses the workplace monitoring.
1.16. Section 10 focuses on the environmental monitoring and radioactive effluent discharge.
Section 11 addresses the personal protective equipment being used by the personnel.
1.17. Sections 12 to 16 are devoted to the control of radioactive material, facility and
equipment design, testing and maintenance of the equipment, radioactive waste management,
transport of radioactive material, and emergency preparedness and response.
1.18. Examples of a safety assessment structure and emergency response procedures can be
found in the Annexes I and II respectively.
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2. JUSTIFICATION OF PRACTICES
2.1. The IAEA Fundamental Safety Principles [9] state that the fundamental safety objective
is to protect people and the environment from harmful effects of ionizing radiation. Principle
4, Justification of facilities and activities, states that “Facilities and activities that give rise to
radiation risks must yield an overall benefit”. This may be taken as equivalent to the wellestablished principle of justification of practices, the operation of radioisotope production
facilities being one example [3].
2.2. The basic requirements for radiation protection for practices established in GSR Part 3
[3] are: justification of practices; individual dose limits; and optimization of protection and
safety.
2.3. When the principle was first formally expressed, many practices, such as the operation
of radioisotope production facilities, were already in widespread use, and in general their
justification was implicit. Under normal conditions, the design, construction, operation and
maintenance of radioisotope production facilities result in doses to workers and the public that
are a small fraction of the respective dose limits in GSR Part 3 [3]. The operation of
radioisotope production facilities can on occasion result in doses to workers and releases of
radioactive material to the environment that may be in excess of authorized limits.
Furthermore, the operation of inadequately designed facilities may result in elevated dose
rates in both uncontrolled areas and unsupervised areas that could result in dose limits being
exceeded. In addition, there are other inherent radiation risks, including those associated with
the security of radioactive material, the transport of radioactive material and also, ultimately,
the disposal of radioactive material.
2.4. IAEA Safety Guide RS-G-1.9 [10] establishes the categorization system of radioactive
sources based on the concept of dangerous quantities of radioactive material (D-values). The
D-value is that quantity of radioactive material, which, if uncontrolled, could result in the
death of an exposed individual or a permanent injury that decreases that person’s quality of
life [11].
2.5. Within this categorization system [10], sources in Category 1 are considered to be the
most dangerous because they can pose a very high risk to human health if not managed safely
and securely. An exposure of only a few minutes to an unshielded Category 1 source may be
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fatal. At the lower end of the categorization system, sources in Category 5 are the least
dangerous; however, even these sources could give rise to doses in excess of the dose limits if
not properly controlled, and therefore need to be kept under appropriate regulatory control.
The finished products of radioisotope production fall into source categories 3–5.
2.6. The decision as to whether the operation of radioisotope production facilities is justified
is specific to the circumstances and benefits of their use, including national priorities, so
definitive recommendations regarding justification cannot be provided. Ultimately, the
decision as to whether the operation of such facilities is justified should be made on a case by
case basis by the appropriate governmental authority, which should weigh the various benefits
and risks associated with their operation in determining whether specific practices are
justified. The governmental authority’s decision as to whether the operation of radioisotope
production facilities in the State is justified may also be made on a general basis for all
radioisotope production facilities of a specific type.
3. TYPES OF RADIOISOTOPE PRODUCTION FACILITIES
3.1. For the purposes of this Safety Guide, general categories of radioisotope production
facilities are defined on the basis of the design of the facility and the resultant radiation
protection provisions necessary:
Facilities that process targets that have been irradiated by a charged particle beam of
an accelerator to produce radioisotopes;
Accelerator facilities with energies of less than 70 MeV/nucleon that are operated
principally to produce radioisotopes. This document addresses these four categories of
accelerators:
o Low energy (<20 MeV /nucleon) cyclotrons for medical radioisotope
production;
o 20–40 MeV/nucleon isotope production cyclotrons;
o >40 MeV /nucleon cyclotrons for mixed research and radioisotope production;
and
o Linear accelerators used for radioisotope production.
When recommendations in this Safety Guide only apply to specific categories of radioisotope
production facilities, those categories are identified.
5
Irradiation of targets in accelerators
3.2.Accelerators for the production of radioisotopes are generally located in the same building
as where the radioisotope containing products are synthesized.
3.3. Accelerators can be used for the activation of isotopes for research and
radiopharmaceutical usage. For the production of
18
F, the target is irradiated and the liquid
mixture (18O-water containing 18F) is transferred in capillary pipes to a processing hot-cell.
3.4.
Accelerators are used for activation of isotopes for research and radiopharmaceutical
usage. Such accelerators are designed and sold to isotope production facilities or hospitals.
Some accelerators are designed specifically for positron emission tomography (PET)
radiopharmaceuticals, e.g.
18
F. Examples of accelerator types I–V can be found in Section 6
of Ref. [12].
4. DUTIES AND RESPONSIBILITIES
GENERAL
4.1. The person or organization responsible for facilities and activities that give rise to
radiation risks should have the prime responsibility for protection and safety. Other parties
should have specified responsibilities for protection and safety. In line with para. 3.6 of the
IAEA Fundamental Safety Principles [9], the operating organization is responsible for:
(a)
Establishing and maintaining the necessary competences;
(b)
Providing adequate training and information;
(c)
Establishing procedures and arrangements to maintain safety under all conditions;
(d)
Verifying appropriate design and the adequate quality of facilities and activities and of
their associated equipment;
(e)
Ensuring the safe control of all radioactive material that is used, produced, stored or
transported;
(f)
Ensuring the safe control of all radioactive waste that is generated and
6
(g)
Establishing plans and procedures to respond to any radiological emergency that may
arise at the facility and coordinating exercises to test the same [13].
4.2. Specific duties and the day to day responsibilities for the design, operation and eventual
decommissioning of the facility will, however, lie with a range of people, including senior
management, the radiation protection officer (RPO), workers who operate the facility and
handle radioactive material, and qualified experts/radiation protection advisers (RPAs).
MANAGEMENT OF RADIATION SAFETY AND SAFETY CULTURE
4.3. The operating organization, through its managers, is responsible for the establishment
and implementation of the technical and organizational measures necessary to ensure
protection and safety and for compliance with the relevant legal and regulatory requirements.
If this expertize is not available in house, an external qualified expert/RPA should be
appointed to take responsibility, for radiation safety and regulatory compliance.
4.4. A senior manager should be designated as having overall responsibility for overseeing
radiation safety, and verifying that all activities involving radioactive material are carried out
in accordance with regulatory requirements. Responsibilities for radiation safety are required
to be established, and they should be agreed to by all relevant parties and recorded in written
form. Managers should ensure that procedures are in place for the protection of workers, the
public and the environment, and for ensuring that doses are kept as low as reasonably
achievable (the principle of optimization). All policies and procedures should be documented,
and should be made available to all staff and the regulatory body as appropriate.
4.5. Managers are required to foster and sustain a strong safety culture within their
organization, to encourage a questioning and learning attitude to protection and safety, and to
discourage complacency with regard to safety [14]. A strong safety culture is promoted by
management arrangements and workers’ attitudes, which interact to foster a safe approach to
the performance of work. Safety culture is not confined to radiation protection; it should also
extend to conventional safety.
4.6. In cases where there is a potential conflict between operational responsibilities, such as
responsibilities for meeting a production schedule and responsibilities for radiation safety,
radiation safety requirements should always take priority.
7
4.7. Operating organizations with a strong safety culture do not assign blame when incidents
occur; they should learn from their mistakes, foster a questioning attitude and seek continuous
improvement in the safety of work processes. For each incident, the question of acceptable
behavior should be answered on a case by case basis and, in some cases, disciplinary
measures may be taken.
4.8. As stated in GSR Part 2 [14], the operating organization is required to establish,
implement, assess and continually improve an integrated management system that defines the
responsibilities of all relevant persons and that details the key radiological requirements for
personnel, equipment and the facility. The management system should be based on national or
international standards [14, 15, 16]. It should incorporate mechanisms for routine internal
inspections and audits, as well as third party audits, as appropriate. The radiation protection
programme should be integrated into the management system.
Facilities and resources
4.9. The operating organization should ensure that suitable safety systems have been
installed and equipment is available to enable work to be carried out safely and in accordance
with regulatory requirements.
Notification and authorization
4.10. An application for a license should contain information that demonstrates the safety of
the practice. Guidance relating to the preparation of an application for the authorization of a
radioisotope production facility, and its subsequent review by the regulatory body is included
in a GS-G-1.5 [17].
4.11. When applying for an authorization, the operating organization should provide the
regulatory body with the appropriate documentary evidence to demonstrate that an adequate
level of radiation safety will be afforded and maintained.
4.12. The documentary evidence necessary to support an authorization request should
include, as a minimum, specific information concerning the:
(a)
Identification of the operating organization and the individual(s) representing the
operating organization;
(b)
Radioisotopes and chemical forms of the material to be possessed;
8
(c)
Characteristics of the particle accelerator, i.e. type (cyclotron, linear accelerator),
energy, current, beam characteristics and layout;
(d)
Facility in which particle accelerators and/or radioactive material will be processed and
stored with particular attention paid to associated safety systems and equipment, e.g.,
radiation shielding, interlock systems, fume hoods, remote handling tools, effluent
exhaust systems, monitoring systems, and warning systems;
(e)
Locations where particle accelerators will be operated and radioactive material will be
processed and stored;
(f)
Inventory system to be used to account for radioactive material;
(g)
Identification and details of qualifications of the radiation protection officer (RPO) and,
where appropriate, qualified experts or radiation protection advisers (RPAs);
(h)
Operating organization’s requirements for the training and qualification of all relevant
staff;
(i)
Justification for the operation of the facility;
(j)
Safety assessment covering the operation of the facility;
(k)
Radiation protection programme;
(l)
Arrangements for the management of radioactive waste; and
(m) Arrangements for responding to a radiological emergency within the facility premises
(see Section 16).
4.13. The operating organization should obtain the approval of the regulatory body before
commencing a new facility or implementing modifications to the facility. The operating
organization should notify the regulatory body of any changes to key personnel, in particular
senior managers, the principal radiation protection officer and qualified experts/radiation
protection advisers.
RADIATION PROTECTION OFFICER
4.14. The operating organization should appoint at least one employee as a radiation
protection officer (RPO) to oversee the day to day implementation of the radiation protection
9
programme and to carry out the duties required by the programme. While the RPO oversees
the application of the safety standards, the prime responsibility for safety remains with the
operating organization. The RPO should be technically competent in radiation protection
matters of relevance for a given type of radioisotope production facility. The RPO should
report directly to senior management and should have sufficient authority to discharge his/her
duties. Where there is a conflict between safety and operations, the RPO should have the
authority to stop work that is at risk.
4.15. During times when the RPO is not available to provide oversight on radiation safety
matters, such as during periods of absence from the facility, arrangements should be made for
the prompt provision of authoritative advice concerning radiation safety matters. Such
arrangements could include timely access to qualified experts/RPAs or the designation of
deputy RPOs who are present at the facility during times of operation.
4.16. The duties of the RPO should include the following, some of which may require
consultation with, or assistance from, a qualified expert:
(a)
Oversight of facility operations to assist the operating organization to comply with
regulatory requirements;
(b)
Optimizing exposure controls and maintaining safety systems and other equipment that
contributes to controlling exposure of workers and members of the public;
(c)
Oversight of the inspection and maintenance of safety systems, radiation monitoring
equipment and warning features;
(d)
Establishment of controlled areas and supervised areas and oversight of access control
for controlled areas;
(e)
Periodic review of arrangements for individual monitoring of workers;
(f)
Investigation of high, unexpected or reportable exposures and overexposures;
(g)
Ensuring that workers are suitably trained in the use of equipment and in radiation
protection, and that they receive regular refresher training;
(h)
Ensuring that emergency plans and procedures are established and maintained and
exercises are conducted as appropriate (see Section 16);
10
(i)
Oversight of arrangements for environmental monitoring, including review of the
results of such monitoring;
(j)
Establishment, issue and periodic review of local rules (including work permits where
appropriate);
(k)
Investigation and reporting of incidents and accidents;
(l)
Liaising with contractors, designers and suppliers with regard to radiation protection
matters and significant changes to physical or operational aspects of the facility;
(m) Ensuring the adequacy of safety assessments and emergency plans for any reasonably
foreseeable incidents with consequences for radiation protection;
(n)
Oversight of issues related to the safe transport of sources including the receipt of
packages containing radioactive material and the preparation of packages for shipment,
and
(o)
Maintaining records relevant to the radiation protection programme including records
concerning the radioactive material inventory, workplace monitoring, individual
monitoring, environmental monitoring and radioactive waste.
QUALIFIED EXPERTS/RADIATION PROTECTION ADVISER
4.17. A qualified expert/RPA is an individual who is duly recognized, by virtue of
certification by appropriate boards or societies, professional licenses or academic
qualifications and experience, as having expertise in a relevant field of specialization. The
qualifications of a qualified expert are described in paras 3.65–3.71 of Ref. [18].
4.18. The operating organization may identify one or more qualified experts/RPA to provide
advice on various matters concerning radiation safety in the design and operation of the
facility. A qualified expert/RPA need not be a full time employee of the operating
organization but could be employed on a part-time or an ad hoc basis. Regardless,
arrangements should be made for the advice of a qualified expert/RPA to be available when
necessary. As with the RPO, the operating organization cannot delegate its responsibility for
safety to a qualified expert.
11
4.19. A qualified expert/RPA should be experienced in radiation protection matters and
should have had:
(a)
Theoretical training that includes training in radiation protection and the properties of
radiation as used in the radioisotope production facility:;
(b)
A thorough knowledge of the hazards associated with the radiation present and the ways
in which the hazards can be controlled and minimized;
(c)
A knowledge of the emergency preparedness category of the facility in the context of
the emergency preparedness and response (EPR) plans conforming to relevant
requirements of the international standards [13];
(d)
An understanding and detailed knowledge of the working practices used in the facility,
as well as general knowledge of the working practices in other similar facilities;
(e)
A detailed working knowledge of all regulatory provisions, relevant codes of practice
and protection standards, guidance material and other information necessary for giving
advice in connection with the work with radiation undertaken by the operating
organization;
(f)
An awareness of regulatory requirements that could affect the work with radiation on
which the qualified expert/RPA gives advice;
(g)
The ability to give advice so that the operating organization can comply with regulatory
requirements and follow good radiation protection practices;
(h)
The personal qualities to be able to communicate with workers and their representatives;
(i)
The ability to keep up-to-date with developments in the use of radiation in the field in
which the qualified expert/RPA gives advice and with developments in radiation
protection.
4.20. The operating organization should provide the qualified expert/RPA with adequate
information and resources as may be necessary for the expert to work effectively. The
information should include a clear statement of the scope of the advice that the expert is
expected to give.
12
4.21. The operating organization may consult the qualified expert/RPA on a wide range of
issues relating to radiation safety, including:
(a)
Optimization of protection and safety;
(b)
Maintenance of engineering features and other equipment;
(c)
Workplace, individual and environmental radiation monitoring;
(d)
Investigation of high exposures and overexposures;
(e)
Staff training;
(f)
Safety assessment and emergency arrangements3;
(g)
Examination of any plans for a facility or for modifications of an existing facility;
(h)
Independent audits related to radiation safety matters;
(i)
Quality management;
(j)
Emergency preparedness and response (see Section 16).
WORKERS
4.22. While the primary responsibility for radiation safety lies with the operating
organization, workers (including assistants and trainees) have a responsibility to work safely
and to take all reasonable actions to restrict their own exposure and those of other workers
and members of the public. Workers include individuals whose work involves exposure to
radiation or work activities that could result in exposures to other individuals or the
environment such as process operators, operators working with product shipment, operators
working with waste, research scientists, pharmacists, laboratory technicians, personnel with
housekeeping duties and personnel who perform routine maintenance activities. The
3
In line with Ref. [13], emergency arrangements are “the integrated set of infrastructural elements, put in place
at the preparedness stage, that are necessary to provide the capability for performing a specified function or task
required in response to a nuclear or radiological emergency” and include: assignment of authorities and
responsibilities, organization, coordination, personnel, plans, procedures, facilities, equipment, training,
exercises, quality management programme etc.
13
competence of these workers to perform their duties in a safe manner should be verified by a
RPO. Workers should:
(a)
Follow the local rules (para 4.26) and any relevant procedures;
(b)
Wear their individual dosimeters in the correct place at all times during radiation work
and record their daily doses. If the dose exceeded the level set by the local rules they
should report it to the responsible (senior) manager or RPO (see Section 6);
(c)
Use radiation monitors properly and in a systematic manner (see Section 8);
(d)
Cooperate with the RPO and qualified experts on all radiation safety issues;
(e)
Participate in any training concerning radiation safety including emergency drills and
exercises;
(f)
Abstain from any willful action that could put themselves or others in contravention of
regulatory requirements or of the operating organization’s own requirements.
4.23. Workers should promptly inform the RPO of any incident or circumstances that could
result in higher than usual radiation doses to themselves or to other persons. This could
include failures or observed deficiencies in safety systems and warning systems, errors in
following procedures, or inappropriate behaviour. A written report should be made to the
RPO as soon as practicable after the incident or observation.
4.24. Radiation safety should be incorporated into the daily routine of work by all personnel.
4.25. Temporary workers should comply with work practices and local rules within the
facility.
LOCAL RULES AND PROCEDURES
4.26. The operating organization should ensure that local rules and procedures are fully
understood by the workers and should, as a minimum, include (see also [18], paras 3.87–
3.92):
(a)
A description of the nature of the hazards posed by the facility and the safety features
used to minimize the risks;
(b)
Written emergency plans, procedures and instructions in line with their respective duties
(see Section 16);
14
(c)
A description of the functions, duties and responsibilities with regard to radiation safety
of key individuals within the operating organization, including the qualified expert/RPA
and RPO;
(d)
The method of ensuring that persons entering controlled areas are wearing appropriate
radiation monitoring devices and that the results of the monitoring are recorded;
(e)
Access and egress monitoring procedure for workers and visitors;
(f)
Written instructions covering actions to be taken in the event of malfunctions. These
instructions should identify individuals to be notified in the event of a malfunction and
should provide a general outline of the corrective actions to be taken;
(g)
Written instructions to ensure that the facility is maintained as prescribed in design
documentation. Written instructions to require that the worker call for assistance from
the RPO when a hot cell or particle accelerator shielding is opened;
(h)
Written instructions describing the wearing of suitable personal protective clothing in
supervisory and controlled areas;
(i)
Written instructions to require that the workers check with the RPO that the plant is safe
before entrance.
5. SAFETY ASSESSMENT
GENERAL
5.1. General safety requirements on safety assessment for facilities and activities are
provided by GSR Part 4 (Rev. 1) [19] which includes the necessary arrangements for:
(a)
scope, purpose, and responsibilities (overall requirements 2–4),
(b)
specific requirements (5–12),
(c)
defence in depth and safety margins (requirement 13),
(d)
safety analysis (requirements 14–19),
(e)
documentation, independent
(requirements 20–24).
verification,
management,
use
and
maintenance
15
5.2. Requirements of GSR Part 4 (Rev. 1) [19] mentioned above in para. 5.1 and
requirement 13 on safety assessment in GSR Part 3 [3] are addressed in this Section in respect
of radioisotope production facilities.
PURPOSE AND DEVELOPMENT PROCEDURE
5.3. Requirement 4 (Purpose of the safety assessment) of GSR Part 4 (Rev. 1) [19] requires
that the primary purposes of the safety assessment is to determine whether an adequate level
of safety has been achieved for a facility or activity, and whether the basic safety objectives
and safety criteria established by the designer, the operating organization and the regulatory
body, in compliance with the requirements for protection and safety as established in GSR
Part 3 [3], have been fulfilled.
RESPONSIBILITY FOR SAFETY ASSESSMENT DEVELOPMENT
5.4. GSR Part 3 [3] states that the person or organization, or registrants and licensees, as
appropriate, is required to conduct a safety assessment that, depending on the type of practice
or source, is either generic or specific to the practice or source for which they are responsible.
5.5. The preparation for the safety assessment, in terms of assembling the expertise, tools
and information required to carry out the work is addressed in requirement 5 of GSR Part 4
(Rev. 1) with a detailed description in paras 4.18 (a)–(d) [19].
5.6. The operating organization of the radioisotope production facility should be responsible
for the fulfilment of requirements mentioned above in paras 5.4–5.5 [3, 19].
5.7. An example schematic of a safety assessment for a radioisotope production facility is
illustrated in Figure 1. This figure outlines the key aspects of the radioisotope production
facility which should be addressed in a safety assessment. Thereafter, each of the individual
risk assessments (e.g. shielding, emissions, engineering controls, decommissioning, etc.)
should be collated into a safety assessment report for the facility. The same approach should
be adopted whether it is for a new standalone facility or a modification to an existing and
approved facility. The relevant requirements for performing the appropriate risk assessments
are provided in para. 5.1. Some specific examples of safety requirements (e.g. shielding,
interlocks, transfer lines, remote handling, fume hood, ventilation etc.) are provided in paras
5.13–5.45.
16
Business/Clinical/Research Requirement
Cyclotron Facility
Risk Assessment/Safety Assessment
Radiation Protection
Site Characteristics
Shielding
Site Management
Engineering Controls
Radiation Monitoring
Emissions
•
•
•
Filtration
Abatement
Modelling
•
•
•
Solid
Liquid
Gas
Waste Management
Design Dose Criteria
Transport/Movement
Feedback
Request for more information
Personnel Monitoring
Decommissioning
Facility upgrade
Safety Report
Approval
Review
Submission
FIG. 1. Example schematic of a safety assessment for a radioisotope production facility.
5.8. Requirement 6 of GSR Part 4 (Rev. 1) [19] states that the possible radiation risks
associated with the facility or activity shall be identified and assessed. An example schematic
of the key radiological risks associated with a radioisotope production facility is presented in
Figure 2.
Accelerator/Cyclotron
Radioisotope
Hot Cell (P1)
Air Pressure Control
(negative)
Emission Control
•
•
•
•
Synthesis
Air Pressure Control
(negative/positive)
Abatement
Filtration
Engineering Controls
Continuous Radiation monitoring
Liquid
Finished Product
GMP
Dispenser Hot Cell (P3)
Product Transfer (P2)
Terminal Filtration
Vial Filling
Packaging
Transportation
P2 > P1 then P3 > 0
P2 ≤ P1 then P3 < 0
FIG. 2. Key radiological risks associated with the isotope production facility.
5.9. During and post irradiation, there is a risk of volatile radioactive products being released
to the environment; this may occur while the product is being transferred to the hot cell.
17
5.10. During synthesis in the hot cell, there is a potential risk of radioactive contamination of
the environment outside and inside the building which could result in potential exposure of
operational staff and a limited number of the members of the public in the local vicinity. This
risk is directly related to the potential presence of volatile products within the hot cell during
radiosynthesis. The risk of this contamination should be minimized by an appropriate
negative pressure regime in the hot cell. The risk of release to the atmosphere should be
controlled by the appropriate engineering controls (e.g. filtration, motorized damper,
abatement system, etc.).
5.11. During filling of the finished product in the dispenser hot cell, the appropriate
engineering controls should be in place (good medical practices (GMP) or other aseptic
standards) to ensure that the operator and the product are protected. Specifically, it is
important that any potential volatile radioactive material in the synthesis hot cell should be
prevented from entering the dispensing hot cell. This can be achieved by using the appropriate
pressure regime or other options (e.g. laminar flow, filtration).4
5.12. Transport of the finished product in the shielded container should comply with the
IAEA Transport Regulations [20].
SAFETY ARRANGEMENTS
Shielding
5.13. Direct radiation exposure of workers and members of the public due to the operation of
radioisotope production facilities should be attenuated to optimized levels by the use of
appropriate shielding. Concrete is often used to construct the radiation room shield, but other
materials such as earth fill, steel and lead may also be used in its construction. The shielding
properties of particular materials are well established [21–28], but experience deriving from
existing radioisotope production facilities should be taken into account. The shielding should
provide adequate reductions in radiation levels to keep doses within the dose constraints
established or agreed to by the regulatory body.
5.14. Penetrations of the shield are necessary for entry and exit ports for personnel and
product and for the ventilation system and other ducting. These penetrations can potentially
4
It is important to note that the dispensing of the finished product for use in humans should comply to local
GMP requirements, which may include dispensing in positive pressure regimes.
18
create particular problems for the shielding designer, who should ensure that there is no direct
radiation leakage path, and should ensure that the use of maze entrances and shield plugs are
sufficient to reduce external radiation fields to optimized levels. Care should be taken to
ensure that all significant radiation paths are fully shielded. Where practical, all tubes, pipes
and conduits should take a curved or stepped path through the shielding material to reduce
external radiation levels or should be embedded in the concrete slab using pits and trenches.
5.15. Once the shield has been designed, no subsequent changes should be made, unless they
have been carefully considered and agreed with the regulatory body.
Remote handling tongs, master/slave manipulators
5.16. Handling of radioactive materials in the hot cells may require a remote handling tool,
such as tongs or robotic manipulators, if the chemical processing system is not fully
automated.
Inner surfaces of hot cells
5.17. The inside of the hot cell should have a leak tight liner that can provide air tightness to
prevent the release of radioactive materials from the hot cells. It should also be suitable for the
process inside of the hot-cell (acid fume resistant coatings in solid target dissolution stations
where hot acid can be present). Edges of the liner should be rounded with an appropriate
radius to prevent the accumulation of contaminated dust. The surface should have
unnecessary protruded parts for easy decontamination of the surface. The liner itself should
have enough mechanical strength to support any heavy system intended to be installed. For
the production of radiopharmaceuticals, the inner liner should be designed to comply with the
air classification and air flow requirements (sometimes in laminar flow when open
radiopharmaceuticals are handled like in filling machines).
Fume hoods
5.18. Fume hoods are appropriate for the handling of hazardous and radioactive materials
when the potential for contamination control is low and when external dose rates are low.
Partial-enclosure fume hoods allow high accessibility by chemists and manipulation of special
equipment while affording protection from chemical fumes and radioactive aerosols. The sash
height should be adjusted to maintain the face velocity (0.5 to 0.75 m s-1) of air entering the
hood opening, which should be greater than the capture velocity of contaminants likely to be
released into the fume hood work area to prevent releases into the general laboratory area.
19
5.19. Fume hoods may require external shielding depending on the dose rate of the intended
operation.
5.20. The inspection/maintenance of the fume hood should be done on a scheduled frequency.
The face velocity should be checked prior to use.5
5.21. The exhaust air should be routed through an appropriate filtration system. This exhaust
air should be monitored based on the concentration of the effluent. The volume of exhaust air
can be determined if the face velocity and sash area is known. The exhaust air should be
routed through an appropriate filtration system to limit releases of radioactive material to
external environments.
Glove boxes
5.22. Glove boxes are air-containment systems that isolate the hazardous or radioactive
materials from the operator’s laboratory environment. Glove boxes can be used for nongamma emitting radioisotopes where the shielding of the hot cell is not required.
5.23. Glove boxes are constructed using mild steel, stainless steel, or aluminium coated on the
interior surfaces with chemical-resistant epoxy paint, laminated safety glass panels for
viewing work activities inside the box, and heavy neoprene gloves (glove port) that allow the
operator to handle materials safely inside the glove box. Glove boxes should be equipped with
adequate lighting.
Clean environment considerations
5.24. In order to maintain a clean environment for the production facility, the production line
should be in a clean room or isolator to ensure the required air classification is achieved. If
cleaning agents are used to achieve a sterile/aseptic environment in the hot cell (e.g. H2O2), a
risk assessment should be carried out to ensure that it does not adversely affect the extraction
filtration system.
5
Fume hoods require a large volume of air and this may have design implications on the volume of air required
in the production facility.
20
5.25. Secondary neutrons generated during radioisotope production give rise to neutron
activation of the cyclotron components and vault room wall. Additional forms of shielding
may be required to attenuate and shield the neutrons.
Interlocks
5.26. An interlock should be installed at the access door to controlled areas such as cyclotron
rooms and target rooms to protect the workers from ionizing radiation.
5.27. Access by personnel to the radiation field following an irradiation, securing of the
radiation room prior to initiating irradiation, and irradiation start procedures should
incorporate a series of sequential safety interlocks and controls. Such safety interlocks and
controls should be so designed that any attempt to pre-empt the controls or to apply them out
of sequence will automatically prevent the intended operation.
Transfer systems
5.28. Transfer systems for the radioactive materials vary depending on what types of
materials are transferred.
5.29. Transferring of radioisotopes from the cyclotron to the hot cell is achieved by using
shielded transfer lines and inert gases to move the product from the target to the hot cell.
5.30. Transferring of the radioactive materials between hot cells can be done either through a
simple shield door and/or a pass box installed between hot cells. Also, a conveyor can be
employed to transfer the radioactive materials. In cases of a liquid, it can be delivered through
the tubing either by vacuum or pressure. Delivery of gases can also be done by using a
method similar to that for liquid. Specially, the gas transfer should be done in a closed system
to ensure that there is no risk of release of radioactivity to the environment.
5.31. Transferring of target materials from the target room to the processing hot cells is
similar to the transfer of radioactive gases and liquids. However, the transfer of solid targets
requires more physical and robust transfer systems, utilising pneumatic systems as opposed to
inert gases.
5.32. Transport of a bulk source, dispensed vials, and sealed sources to outside of the building
should follow the protocol for the transport of radioactive material described in section 15.
21
Ventilation
5.33. For facilities within larger organizations (for example production sites within a hospital
environment) systems/procedures should be put in place to ensure that no technical personnel
can access the ventilation system or power distribution cabinet of the facility without prior
information and consent of the facility management and the radiation safety officer. The
operating organisation needs to enforce appropriate standard operating procedures (SOPs) for
the maintenance of all shared and interfacing infrastructure.
5.34. Air pressure within radioisotope production facilities should generally be kept lower
than the external air pressure at all times so that air flows from outside the facility to the
inside. Any air that leaves the building should pass through ducting equipped with filtration
and monitoring equipment.
5.35. Redundancy of critical ventilators should be in place to:
(i)
Ensure the safety of the site during ventilation maintenance.
(ii)
Ensure back-up power for critical ventilation systems.
5.36. Redundancy of power to critical parts of the ventilation should be in place. Use of
diesel/gas generators and Uninterruptible Power Supply (UPS) should be considered.
5.37. Appropriate filters should be in place for:
(a)
incoming air,
(b)
outgoing air.
This is largely dependent of the chemical compounds produced and the nuclides used. The
filter selection needs to be appropriate for the products being used. Appropriate measures to
contain gases which cannot be trapped by filters should be put in place.
5.38. The ducting (piping) for the intake/exhaust air should be constructed of stainless
steel/mild steel epoxy lined or galvanize: designed as per industry standard.
5.39. The supply air to all ‘clean rooms’ should have terminal HEPA filters which should be
tested as per industry standards. The air handling units should have appropriate intake filters
and set up to condition the supply air. These air handlers should supply 100% fresh air with
no recirculation. The exhaust air should be monitored for radioactive contamination (exhaust
stack) prior to leaving the facility.
22
SITE SELECTION
5.40. During the processes of site selection and site evaluation, particular consideration
should be given to potential hazards that cannot be addressed by means of engineering
measures, such as hazards relating to flooding and hazards relating to geological phenomena
in areas of potential or actual subsidence, uplift, collapse, faulting, volcanic activity [29],
hurricanes, tornadoes and tsunamis.
Safety assessment of waste management
5.41. A safety assessment of waste management at the production facility should be
documented and periodically updated as required. Measures to control the generation of
radioactive waste, in terms of type, volume and activity, should be considered throughout the
lifetime of a facility, beginning with the design phase, through the selection of materials for
the construction of the facility, and by the control of materials and the selection of the
processes, equipment and procedures used throughout operation and decommissioning of the
facility, including:
(i)
Potentially radioactive liquid waste handling system with liquid waste decay tank and
chemical waste from QC operation or target processing (solid target dissolution);
(ii)
Solid waste containment and storage room;
(iii) Gaseous waste; and
(iv) Evaluation of national procedures and availability of a long-term storage facility for
solid waste.
Safety assessment report
5.42. The operating organization should demonstrate to the regulatory body how the design of
the radioisotope production facility and the related operational procedures will contribute to
radiation safety during normal operation, to the prevention of accidents, and to the mitigation
of the radiological consequences of such accidents if they were to occur. This information
should be provided in the form of a documented safety assessment report describing and
evaluating the predicted response of the plant to incidents (including postulated malfunctions
or failures of equipment, common cause failures and human errors) and external events of
natural origin and human induced origin that could lead to accident conditions. These
23
analyses should include the consideration of combinations of such malfunctions, failures,
errors and external events.
5.43. The results of all the risk assessments referred to in this section should be included in
the safety assessment report.
Facility and Equipment Design Specification
5.44. An integral part of the safety assessment is the design specification of the facility and
the equipment to be utilised therein. Each facility design will be unique to the user
requirement specifications, the proposed site and the local regulatory requirements. In
addition to the specific design requirements referred to in the previous section on safety
assessment, some of the key issues which should also be considered when setting up a new
radioisotope production facility or modifying an existing production facility are listed in
Annex I of this document.
6. RADIATION PROTECTION PROGRAMME
GENERAL
6.1. The general objective of a radiation protection programme is to discharge the
management’s responsibility for radiation protection and safety through the adoption of
management structures, policies, procedures and organizational arrangements that are
commensurate with the nature and extent of the radiation risks. The radiation protection
programme represents the totality of actions undertaken to achieve the declared aims of the
operating organization for radiation protection and safety. The radiation protection
programme is a key factor in relation to the development and maintenance of the safety
culture within an organization [18], and it should meet the regulatory requirements. The
operating organization should always strive to minimize doses to workers and members of the
public.
6.2. The operating organization should develop, document and implement a radiation
protection programme [18]. This should include information on the radiation protection
arrangements, the safety assessment, the measures for implementing the arrangements, and
the mechanism for the review and updating of the arrangements.
24
6.3. Application of the optimization principle should be the principal driving force behind
the establishment and implementation of radiation protection programmes, including in many
cases measures to prevent or reduce potential exposures and to mitigate the consequences of
accidents if they were to occur. The existence of a radiation protection programme is not
sufficient in itself; managers and workers should demonstrate their on-going commitment to
the programme and its objectives. Detailed guidance for establishing and maintaining a
radiation protection programme that focuses on the protection of workers is provided in an
IAEA Safety Guide [18].
6.4. The programme should be based on the operating organization’s safety assessment, and
it should address planned exposure situations.
6.5. The operating organization should ensure that information on both normal and abnormal
operations that are relevant to radiation protection and safety be disseminated or made
available, as appropriate, to the regulatory body and to manufacturers or suppliers, as
specified by the regulatory body [3]. Such information should include maintenance data,
descriptions of events, information regarding defects in materials and equipment, weaknesses
in operating procedures, corrective actions, etc. The operating organization should ensure that
any new information of this type that is known to manufacturers and suppliers of equipment is
obtained from them once it is available. It may be necessary for the operating organization to
seek this information from the manufacturer or supplier periodically rather than relying upon
them to provide it.
STRUCTURE OF THE RADIATION PROTECTION PROGRAMME
6.6. The guidance on the radiation protection programme is provided in Section 3 of
Ref. [18]. The radiation protection programme should include a top level policy document
supported by detailed and specific procedures or ‘local rules’ and a comprehensive system of
records (quality management system).
MANAGEMENT STRUCTURE AND POLICIES
6.7. The radiation protection programme should include a description of the management
structure as it relates to radiation safety. This structure, which may be presented in the form of
an organizational chart, should show the names of the senior managers responsible for
radiation safety and of the various duty holders (e.g. the RPO). The chart should clearly show
the line of reporting, from the worker through to the senior manager with overall
25
responsibility. If the operating organization has more than one location of operations, the
management structure should clearly specify the responsible persons at each location.
6.8. The radiation protection programme should include a commitment by the management
to keeping radiation doses as low as reasonably achievable and to fostering a strong safety
culture.
Assignment of responsibilities for radiation safety
6.9. All posts for which responsibilities are allocated should include the senior managers of
the operating organization (which has the prime responsibility for safety), the RPO, the
qualified expert/RPA and other workers who have responsibility towards radiation safety, as
described in Section 2. Personnel must be informed of their responsibility towards radiation
safety. Specific responsibilities towards certain procedures and records should be allocated to
specific workers.
Local rules and supervision
6.10. Local rules that describe the procedures for carrying out radiation work should be
developed and written in a language known by the people who will follow them. These local
rules should cover all procedures associated with work where there is the potential for
radiation exposure, such as routine operations, cell maintenance and transport (see Sections
10 and 11). The local rules are an important tool in the restriction of radiation doses. They
should include sufficient information and guidance to allow workers to carry out their duties
safely and in compliance with regulatory requirements.
6.11. Management should ensure that all relevant persons have read and understood the local
rules. A copy should be provided to all workers and other relevant persons, and additional
copies should be available in the work area. In smaller organizations with a limited amount of
work, it may be appropriate to have one set of local rules covering all procedures.
6.12. In larger organizations, it might be appropriate to have several sets of specific local
rules. A facility specific procedure should be established. Workers should be informed on
such procedures.
6.13. A short version of the local rules should be approved for visitors to review and
understand.
26
6.14. Visiting workers should be made aware and trained in relevant sections of the local
rules.
6.15. The operating organization should appoint at least one employee as a RPO to oversee
the day to day implementation of the radiation protection programme and to carry out duties
as required by the programme. Details of the duties of the RPO are given in Section 4.
6.16. Operating organizations should ensure that female employees who enter controlled or
supervised areas are provided with information regarding the risks to an embryo or foetus
from exposure to radiation and the importance of notifying their employer as soon as
pregnancy is suspected.
Designation of controlled areas or supervised areas
6.17. The radiation protection programme should describe how controlled areas 6 and
supervised areas7 are to be designated for the isotope production facility. Controlled areas
should be established with the goal of restricting exposures of workers in isotope production
facilities. The designation of such areas should be based on the safety assessment.
6.18. The area at the side of the cells where transfer containers are coupled should be
designated as a controlled area. The front of the cell should be designated a supervised area
because there is a lower probability of contamination and radiation. The internal compartment
of all hot cells should be designated as controlled areas.
6.19. The active maintenance area at the side of the cells where transfer containers are
coupled should be designated as a controlled area because of the higher probability of
contamination and radiation in that area.
6.20. The area where the products are received into the hot cell and dispensed normally have
a higher probability for contamination and radiation and should therefore be a controlled area.
6
A controlled area is a defined area in which specific protection measures and safety provisions are or could be
required for: (a) controlling normal exposures or preventing the spread of contamination during normal working
conditions; and (b) preventing or limiting the extent of potential exposures.
7
A supervised area is a defined area not designated a controlled area but for which occupational exposure
conditions are kept under review, even though no specific protective measures or safety provisions are normally
needed.
27
6.21. In the accelerator room there should be low probability of contamination and radiation
and therefore can be operated as a supervised area.
Periodic reviews and audits of the performance of the radiation protection programme
6.22. As an integral part of the operating organization’s management system, the radiation
protection programme and its implementation should be assessed on a regular basis. This
periodic review should identify problems to be addressed and any modifications that could
improve the effectiveness of the radiation protection programme.
6.23. A key part of this periodic review process is a routine series of workplace audits,
including the description and qualifications of the persons who will conduct them, their
frequency, the expectations of the audit team, and the reporting of results and their follow-up.
Quality assurance and process improvement
6.24. Radioisotope production work and its associated activities should be carried out in
accordance with the established management system. This management system should be
designed to ensure that all equipment and safety systems are regularly checked and tested, and
that any faults or deficiencies are brought to the attention of the management and are
promptly remedied.
6.25. The management should also ensure that the correct operational procedures are being
followed, and that the quality assurance programme specifies the relevant checks and audits to
be made and the records to be kept. The relevant regulatory requirements should be taken into
account and reflected in the content and details of the quality assurance programme.
6.26. The management system should include a mechanism for the collection and feedback of
lessons learned from day to day operations, emergencies and incidents (including those
reported both within the organization and in external reports), and how these lessons can be
used to enhance safety.
HEALTH SURVEILLANCE PROGRAMME
6.27. The radiation protection programme should include details of a programme for periodic
health surveillance of radionuclide production personnel and other employees as appropriate.
This should include a requirement to assess the initial and continuing fitness of workers for
their intended tasks. A qualified expert/RPA and/or an appropriately qualified medical doctor
28
should be consulted in the drawing up of the programme for health surveillance, and it should
be consistent with regulatory requirements. No special medical/health surveillance
programme is necessary relating to routine work at an isotope production facility.
RADIATION SAFETY COMMITTEE
6.28. A radiation safety committee should be established for the purpose of regularly
reviewing the performance of the radiation protection programme. In a hospital, the radiation
safety committee for the hospital should also look at the radiological safety aspects of the
cyclotron. In the case where the hospital doesn’t have a radiation safety committee such a
committee should be established. This committee may be dedicated to radiation safety or it
may have other (conventional) safety related responsibilities. The committee should include
the senior manager(s) responsible for radiation safety, the RPO(s), qualified experts/RPA and
representatives of the workforce. The responsibilities of the radiation safety committee should
include, but not be limited to:
(a)
Regular reviews of all aspects of the radiation protection programme;
(b)
Reviews of occupational radiation doses and any accident reports prepared by the RPO;
(c)
Making recommendations for improvements in the radiation protection programme;
(d)
Provision of guidance and direction on the performance of the RPO’s duties;
(e)
Preparation and dissemination of regular reports to all staff about relevant radiation
safety issues.
7. TRAINING AND EDUCATION
GENERAL
7.1. Persons performing work in controlled areas within an isotope production facility are
responsible for ensuring that their work is carried out safely and in compliance with all
relevant regulations and safety standards. Operating organizations should, therefore, ensure
that radiation work is carried out only by workers who are trained, and who are competent and
trained in radiation protection and safety.
29
7.2. The workers in isotope production facilities should have training and qualifications that
are specifically related to their area of responsibility. Some of this training may include only a
limited amount of training in radiation protection and safety. In this case, they should be
supplemented with additional training specifically in radiation protection and safety. Such
additional training may be provided by specialized training organizations rather than by the
operating organization.
7.3. Emergency workers should be qualified and trained in arrangements for preparedness
and response for an emergency that can arise in the course of the production, use or transfer of
radionuclides (see Section 16).
TRAINING PROGRAMME
7.4. The radiation protection programme should describe the full scope of the training
programme in radiation protection and safety for all employees directly involved in routine
isotope production activities and emergency response. It should include a radiation
‘awareness’ programme, where appropriate, for other staff, including managers, research
scientists, laboratory technicians, trainees, workers such as cleaners and maintenance staff
who may be inadvertently exposed, and contractors. The radiation protection programme
should also specify the minimum educational and professional qualifications for all relevant
staff including those involved in an emergency response, especially the RPO, hot cell or
cyclotron operators, and pharmacists in accordance with regulatory requirements.
7.5. The requirements for keeping training records should be consistent with regulatory
requirements and recommendations, and they should be specified in the radiation protection
programme.
7.6. The training programme should be reviewed periodically or when there are significant
changes in design of the facility or processes.
Design of a training programme
7.7. The operating organization should define necessary competences and knowledge for
operating the facility.. This training programme in radiation protection and safety may be
provided by the operating organization or by a specialized training organization. The
operating organization should take into consideration the levels of competence on the basis of
the workers’ training and experience. In the case where an operating organization does not
have the capability or resources to establish a training programme, the workers should attend
30
a training programme on radiation protection and safety provided by competent training
providers, including post-secondary education institutions, radiation protection institutions
and training consultants.
7.8. Programmes should be established for the different levels of training corresponding to
the responsibilities of the worker. The workers could be divided into the following groups:
Hot cell and cyclotron operators
Pharmacists
Radiation protection officers
Laboratory technicians
Research scientists
Maintenance personnel, packaging personnel and decontamination workers.
7.9. The training programme should establish the criteria for passing theoretical and
practical examinations, as well as the procedures to be followed if an applicant fails an
examination. The details of the training programme should be incorporated into the radiation
protection programme.
STRUCTURE AND CONTENT OF THE TRAINING COURSE
7.10. Each training course should be structured around specific aims and objectives and
should be customized to the needs of the target audience. Fundamental concepts and
measurements include:
— Basic ionizing radiation concepts;
— Ionizing radiation quantities and units;
— Ionizing radiation detecting instruments;
— Biological effects of radiation;
— System of radiation protection (radiation protection principles of justification,
optimization and dose limitation);
— Regulatory requirements;
— Designation of controlled areas and of supervised areas;
— Dose limits, dose constraints and investigation levels;
31
— Effects of time, distance and shielding;
— Individual monitoring, external and internal monitoring and how to interpret their
doses;
— Working practices to limit doses and maintain them as low as reasonably achievable;
— Radiation protection programme;
— Emergency preparedness and response.
Practical radiation protection
— handling of radioactive materials including those in unsealed forms;
— implementation of emergency arrangements;
— Specific task related issues.
Hot cell operators
— Operation of hot cell (opening hot-cells for operation or maintenance etc.);
— Manipulator handling (e.g. tongs).
Research scientists
— Specific training on radiation protection without standard working procedures.
Maintenance services
— Maintenance on target system, isotope transfer system, hot cells and manipulators and
operations significant to radiation safety.
Decontamination services
— Decontamination after radioactive contamination incidents.
Waste operators
— Handling instructions for radioactive waste.
Shipping clerks
— IATA training on shipment of radioactive material;
32
— Storage of radioactive materials;
— Access control procedures;
— Security procedures;
— Local rules;
— Management of radiation protection;
— Transport of radioactive materials;
— Measurement of radiation fields and the units of measurement;
— Accidents and other incidents involving the production, use and transport of
radioisotope, their consequences and lessons learned.
7.11. The training should provide practical exercises, including the rehearsal of dealing with
abnormal events (e.g. a broken vial with a medical isotope during dispensing). However,
radioactive sources, unless they are exempt, should never be used in such rehearsals. Not-inuse cells can also be used for training in the use of manipulators and coupling and uncoupling
of transfer containers.
7.12. A RPO and a qualified expert/RPA should provide advice on staff training needs and on
how those needs may best be satisfied. In many cases, a RPO should be able to provide much
of the necessary training.
7.13. Where appropriate, workers should receive adequate training and refresher training in
the proper use of personal protective equipment.
REFRESHER TRAINING
7.14. Management should ensure that their workers’ knowledge and skills are kept up to date
through a programme of refresher training. Such training should include a review of the
fundamentals of protection and safety, and information on changes to equipment, policies and
procedures, and possible changes in regulatory requirements.
7.15. The frequency of refresher training should be consistent with regulatory requirements.
Refresher training is typically given at intervals of less than two years but not exceeding five
years. However, changes in regulations or notifications of safety issues should be
disseminated as written instructions as soon as practicable, and then followed up by inclusion
in refresher training.
33
8. INDIVIDUAL MONITORING OF WORKERS
INDIVIDUAL DOSE ASSESSMENT AND RECORD KEEPING
8.1. Production of radioisotopes increases the potential for exposure to ionizing radiation,
radioactive substances and aerosols by workers. External ionizing radiation fields are created
during the process of target irradiation.
8.2. For radiation safety and regulatory compliance, all workers at a radioisotope production
facility with potential for radiation exposure in controlled areas should be monitored to assess
external and internal radiation dose.
8.3. Target assemblies are encapsulated to limit the release of radioactive material or
aerosols to the work environment. However, work activities during radioisotope production,
target processing, radiochemical separation and purification activities, and radioisotope
handling and packaging activities increase the potential for release and inadvertent intakes of
radionuclides by workers. For work activities having increased potential for internal exposure,
workers should be monitored by direct measurements and indirect bioassay to assess internal
intake of radioisotopes [18].
8.4. The designation of controlled and supervised areas should be reviewed regularly, and
may be changed or extended during initial installation, maintenance, and in order to meet the
operational requirements of the facility.
8.5. All visitors should be supplied with individual dosimeters. The isotope production
facility should record the dose received by the visitor.
8.6. Dosimetry records provide the means for tracking individual radiation exposures and
internal dose from sources of ionizing radiation for both routine work and inadvertent or
accidental exposures. Radiation dose records should be used to demonstrate regulatory
compliance and support radiation safety planning. These records should include the results of
individual worker monitoring for both external radiation and intakes of radioactive material.
Records should include all applicable measurement data, measurement dates and times, names
of personnel monitored individually, and methods used to measure external dose or calculate
internal dose. Personal exposure and dosimetry records should be permanently maintained in
retrievable forms.
34
8.7. States should establish a national dose register for workers in order to accumulate all
doses workers received at different facilities.
EXTERNAL DOSIMETRY
8.8. Individual monitoring tracks individual cumulative exposure, demonstrates the current
level of the occupational radiation safety at a radioisotope production facility and provides
essential information for record keeping. Guidance for establishing external radiation
monitoring for individual workers is given in [18].
8.9.Workers who enter controlled areas in radioisotope production facilities should be
monitored continuously for exposure to ionizing radiation using appropriate methods and
technology.
8.10. A programme for individual monitoring of external radiation exposure is intended to
demonstrate that workers’ exposures are being monitored, to provide information for the
optimization of protection and to verify the adequacy of work procedures. Guidance on
establishing monitoring programmes for external exposure, selection of appropriate
dosimeters, interpretation of results, record keeping and quality management is given in [18].
Types of external monitoring
8.11. Each worker should wear an above-the-waist, whole-body dosimeter (film,
thermoluminescent chip, or optically stimulated luminescence crystal) capable of accurately
recording and integrating cumulative exposure to gamma radiation.
8.12. Hot cell operators, RPOs, pharmacists, decontamination workers, laboratory technicians
and maintenance staff who routinely enter controlled areas should be subject to individual
dose monitoring. These individuals should wear whole body monitors (e.g. a film badge,
thermoluminescent dosimeter or optically stimulated luminescent dosimeter) and an electronic
personal dosimeter to ensure effective dose management.
8.13. Workers who handle or process beta-emitters in close proximity to the eyes and skin
surfaces should wear multi-purpose (gamma, beta) dosimeters with capability for thinwindow beta-ray detection.
8.14. Finger rings should be worn for situations requiring the monitoring of exposure to the
hands.
35
8.15. Eye dosimeters should be worn on forehead for situations requiring the monitoring of
the eye doses. In some cases, it might not be possible to wear eye dosimeters on the forehead
[30].
8.16. The worker should wear the dosimeters under the lab coat, apron or overalls in order to
reflect the dose to the body. It will also prevent the radioactive contamination of the
dosimeter.
8.17. The dosimeters should be processed (or evaluated or read) at least quarterly or more
frequent depending on the nature of work and technical specification of the dosimeter.
8.18. The electronic dosimeters should be used in an isotope production environment
whenever multiple or variable work activities are performed, such as equipment maintenance
or hot cell modifications, involving potentially hazardous radiation levels.
8.19. The tools and procedures for individual monitoring for exposure of workers, including
the type of dosimeter required and the necessary frequency of replacement, should be chosen
in consultation with a RPO or qualified expert or RPA, in accordance with the requirements
of the regulatory body. The dosimeters should be provided and processed by a laboratory or
company that has been authorized by the regulatory body and is traceable to a standards
dosimetry laboratory approved by the regulatory body.
8.20. The operating organization should make arrangements to ensure that dose records are
maintained for each worker in the manner specified in regulatory requirements. Operating
organizations should ensure that personal dose records are provided to workers upon
termination of their employment and are available to the individual at other times.
8.21. Operating organizations should prepare procedures describing the way in which
individual dosimeters are to be administered, and these procedures should include the
following:
(i)
Ordering and receiving dosimeters from the dosimetry laboratory;
(ii)
Distribution of dosimeters to monitored workers;
(iii) Collection and dispatch of dosimeters to the dosimetry laboratory for processing;
(iv) Review and maintenance of dose records.
36
8.22. Operating organizations should provide suitable storage facilities for personal
dosimeters not in use that protect the dosimeters from inadvertent exposure to radiation and
from adverse environmental conditions such as extremes of heat or cold and/or humidity.
Personal dosimeters should not be stored close to any area where dose rates are above normal
background levels. Dosimeters should not be put through mail inspection systems that utilize
X rays.
8.23. Monitored workers should be required to take good care of their dosimeters, and to take
precautions to protect them from loss, theft, tampering or damage and from inadvertent
exposure to radiation. Workers should return dosimeters promptly at the end of the specified
period of wearing. Workers should inform a RPO without delay if a dosimeter is missing or
damaged or if it has been exposed to radiation when they were not wearing it.
8.24. If a dosimeter is lost, all reasonable steps should be taken to recover it. If the dosimeter
cannot be located, operating organizations should carry out an investigation and should
prepare a report that includes an estimate of the dose received by the worker for the relevant
period of time. In some States, the approval of regulatory bodies may be required prior to the
entry of such estimates into a person’s dose record.
INTERNAL DOSIMETRY
8.25. The probability for internal intake of radioactive substances should be established
during the safety assessment of the isotope production facility. A monitoring programme
should be established in cases where there is a probability of intake. The frequency of the
monitoring and the type of monitoring should be determined from the level of probability for
the intake. Guidance on internal dosimetry is established in Ref. [18].
Types of internal dosimetry
8.26. Methods for assessment of radioisotope intakes include direct in vivo counting,
bioassay measurements of urine, faeces, sputum, nasal swipes, or blood, and biokinetic
modeling using measurement data and information on the chemical and physical
characteristics of the material to which workers may be exposed.
8.27. Methods used to assess radioactivity intakes and uptakes should be appropriate for the
radioisotopes under consideration, e.g. for beta emitters a 24-hour urine sample should be
taken and sent for analysis for the isotope in the urine. From the result, internal doses should
37
be calculated for intakes of radioactive materials by workers at isotope production and
processing facilities.
8.28. Biokinetic models have been developed for a broad array of radioactive material forms,
modes of intake, and metabolic pathways to facilitate calculation of internal dose to the whole
body, critical organs, and tissues. Internal dose calculations are facilitated using computer
software or dose-conversion factors per unit intake.
Criteria for internal monitoring
8.29. Under normal conditions the contamination level in the air should be <1/10 DAC
(derived air concentrations) of the isotope 131I. DAC values are available in the IAEA Safety
Guide on occupational exposure [18].
8.30. In cases where there is a probability that contamination in the air could exceed 1/10 of
DAC of the applicable isotope, a routine internal monitoring programme should be
established for the workers that would be appropriate for this isotope.
INVESTIGATION OF DOSES
8.31. The operating organization should instruct workers to notify a RPO immediately if they
know or suspect that they have been exposed to high level radiation or airborne
contamination. If the individual(s) concerned was wearing a personal dosimeter, it should be
sent immediately to the dosimetry laboratory and the laboratory should be informed of the
urgency of the case. In the case of exposure to airborne contamination, the person should be
monitored for the appropriate isotope.
8.32. The operating organization should conduct a formal investigation, as required by the
regulatory body, whenever the recorded dose exceeds the investigation level. The
investigation should be initiated as soon as possible following the event, and a written report
should be prepared concerning its cause. This report should include a determination or
verification of any doses received, details of corrective or mitigating actions, and instructions
or recommendations on how to avoid a recurrence.
8.33. The report should be provided to all concerned parties within the appropriate time frame
as required by the regulatory body.
38
9. WORKPLACE MONITORING
9.1. Paragraph 3.96 of GSR Part 3 [3] states that registrants and licensees, in cooperation
with employers where appropriate, shall establish, maintain and keep under review a
programme for workplace monitoring under the supervision of a RPO or qualified
expert/RPA.
9.2. The type and frequency of workplace monitoring shall:
(a)
Be sufficient to enable:
(i) Evaluation of the radiological conditions in all workplaces;
(ii) Assessment of exposures in controlled areas and supervised areas;
(iii) Review of the classification of controlled areas and supervised areas;
(b)
Be based on dose rate, activity concentration in air and surface contamination, and their
expected fluctuations, and on the likelihood and magnitude of exposures in anticipated
operational occurrences and accident conditions.”
9.3. GSR Part 3 should be consulted for further details on general requirements. Production
facilities have some specific requirements regarding workplace monitoring on fixed and
portable radiation dose rate meters, contamination control and air sampling.
9.4. Dosimetry should be performed by calibrated and suitable instrumentation. Details of
the guidance on the selection of the proper radiation survey instrument for a given application
are provided in Refs [17, 18]. The following subsection summarizes information in regard to
the radiation meters and monitors that are normally employed in the radioisotope production
facilities.
RADIATION METERS AND MONITORS
Fixed and Portable Radiation Dose Rate Meters
9.5. For both fixed and portable dose rate monitors, the detector probes and detector
windows should be carefully selected to suit the type of radiation being emitted. Under
production conditions in the hot cell it is possible to measure beta emitting product at the
outlet of the hot cell after the end of the technology process. It is not often practicable to
39
measure beta radiation inside the hot cell because of the presence of mixed gamma and beta
radiation. Depending on the activities in the production facility, a range of radiation detectors
may be required.
9.6. Fixed radiation dose rate meters are normally referred to as area monitors. Area
monitors serve as an important safety feature to ensure the safety of workers in the workplace.
Alarm levels are set to alert the worker of an elevated radiation dose rate. Both audible and
visual alarm signals should be available to warn personnel on the abnormal situation in the
monitored area. The requirement for the number and location of fixed radiation dose rate
meters should be based upon the safety assessment. Locations for fixed dose rate monitors can
include:
(a)
door openings from hot cells, cyclotron bunkers and caves, with a probe inside the
enclosure interlocked to the door control;
(b)
locations where maintenance activities may inadvertently cause elevated dose rates, for
example at the front of hot cells, shielding covering filtration, ventilation plant room,
waste room, etc.
9.7. A final consideration in determining the location and alarm presets of area monitors is
the avoidance of nuisance alarms. In a production facility, loaded packages and raw materials
are in movement throughout the site, so it is important to have area monitors set not to alarm
due to such routine processes. Routine operational verifications should be preceded by a
verbal alert that testing is happening. If these steps are not followed, workers may not be
alerted to an irregular condition.
Considerations for Portable Dose Rate Meters
9.8. Since ionizing radiation cannot be noticed by the human senses, work at production
facilities can only proceed if trained people have working detection equipment. Therefore, the
number of portable detectors in a production facility is a crucial matter. The production
facility may need a variety of different dose rate meters, for example:
Large volume, thin end window open air ion chambers for beta, and low energy
gamma (<~50 keV) dose rate evaluation (chambers that have desiccants inside are
important considerations as humidity fluctuations may render the chamber
inoperable). These detectors are useful for obtaining a reliable dose rate at 1 meter for
transport measurements, however, because of their size, they are difficult to use to
40
evaluate contact readings or small diameter beams. High dose rate (smaller volume)
open air ion chambers with thick side walls are very useful at localizing high energy
beta activity/contamination in production hot cells.
Large volume pressurized ion chambers, though they are not capable of beta or low
energy gamma detection, are useful for providing stable dose rate measurements and
do not suffer from humidity fluctuations as they must be sealed in order to maintain
their pressurized gas. These detectors are useful for obtaining a reliable dose rate at
1 metre for transport measurements, however, because of their size, they are difficult
to use to evaluate contact readings or small diameter beams.
Proportional detectors may be used as dose rate meters, though they are more
commonly designed for use as contamination meters. When used as dose rate probes,
proportional detectors will normally be sealed and not suffer the effects of humidity.
Geiger-Mueller (GM) type detectors are available in a variety of sizes and
configurations. Larger probes have increased dead times and are not suited to higher
dose rate measurements, smaller volume probes are best suited for big dose rates and
in evaluating dose rates produced by small diameter beams. GM probes smaller than
an ion chamber provide better evaluation for the dose rates near contact on surfaces.
Thin end window GM probes may be suited for beta detection, though they typically
over respond to low energy gamma rays via the thin window. Thin end window GM
probes often have greater directional dependence than other detectors, which is an
important consideration in training of staff in their use. GM probes are sealed and so
do not suffer from humidity fluctuations. They are the most commonly used detector
type because of their cost, ruggedness and ease of use – but are not best suited to all
types of radiation.
A useful type of portable dose meter in a production facility is one that has an
extending pole. Distance can be maximized using an extender type detector to protect
the employee when encountering high or unknown dose rates. Extending detectors are
essential tools at many production facilities and are used to assess cyclotron and target
interventions, dose rates around duct work and hot cells, and for routine surveys. The
information gathered by an extended detector will inform radiation safety staff
whether it is safe to proceed with work at a closer distance and will be able to estimate
the length of time permissible to perform the planned work.
41
Surface Contamination Detection in the production premises
9.9. Contamination surveys can sometimes be performed using direct measurement but
when there are varying or elevated radiation backgrounds in the production facility they are
more frequently performed by swiping. Routine surveys include checks of equipment and
personnel at barrier doors, and routine floor and surface checks. Minimal frequencies for
routine floor and surface checks should be defined by the operating organization, but the
practice should be commensurate with the risks at the production facility and may vary from
weekly at a small facility, to daily or multiple times a day at large production facilities.
9.10. Contamination monitoring should be performed when utilising glove-boxes and fume
hoods or when non-routine work is being carried out.
9.11. Surface contamination surveys fall into two categories at a production facility: routine
and as needed. Contamination surveys can sometimes be performed using direct measurement
but when there are varying or elevated backgrounds in the production facility they are more
frequently performed by swiping or other means. Routine contamination survey frequencies
and acceptable activity/unit area (Bq/cm2) criteria are defined by the radiation protection
programme. If necessary these values can be conservatively converted to units in which the
detector reports (cps or cpm) for ease of use by the operator. Factors to consider when
stabling these values are swipe efficiency, efficiency of the contamination meter at the swipe
to detector distance it will be used (geometry + detector efficiency for the radioisotope), and
the length of time that is required for the reading to stabilize.
9.12. It is normal practice to assume that 10% of loose contamination is removed on a swipe.
This value can be used in calculations to demonstrate compliance for indirect contamination
surveys.
9.13. Routine contamination surveys are an essential part of the defence in depth concept.
Routine surveys include checks of equipment and personnel at barrier doors, and routine floor
and surface checks. Minimal frequencies for routine floor and surface checks may be defined
by the regulator, but the practice should be commensurate with the risks at the production
facility and may vary from weekly at a small facility, to daily or multiple times a day at large
production facilities. Routine floor surveys in general areas and hallways provide an
indication if contamination is being tracked from processing areas. Indirect floor surveys can
42
be performed by swiping with a dry mop with replaceable cloth and direct checking the mop
for contamination ea.
9.14. As needed, contamination surveys happen whenever processing work is performed,
such as when items enter or exit, cells, glove-boxes and fume hoods or when the potential to
perform intervention work is being evaluated in areas which may have non-fixed
contamination (cyclotron bunkers and caves, cell, etc. work) and also on packages which are
being prepared for shipment.
Room Air Contamination Monitoring
9.15. Typically, there are two methods to assess air concentration in production facilities:
either by a fixed/portable continuous air monitor with a shielded contaminant probe (CAM) or
by performing a grab sample on a filter then removing the filter media for measurement at
another location.
9.16. Grab sample filters can be fixed or mobile. Achieving a flow rate across a filter at the
assumed breathing rate of a worker (for example 20 L/min) normally requires equipment that
is too heavy for a worker to wear. Personal air samplers (PAS) which can be worn on a
worker’s lapel normally operate at low flow rates (for example 2 L/min), are more directly
placed in the workers breathing zone, but may get covered by the workers clothing or have
issues with battery life.
9.17. The following should be considered when establishing the breathing air monitoring
programme:
9.18. Set levels at which a room may not be entered or respiratory protection must or may be
used based upon filter efficiency, detector efficiency, line losses, pump flow rate and dose
conversion factors [31] for inhalation.
9.19. Place alarming CAMs in locations of high risk (radioiodine processing areas, waste,
cyclotron bunker and caves) and have the alarm register at a secondary location.
9.20. The number of bends in tubing for CAMs needs to be minimized to avoid line losses.
Tubing material for CAMs needs to be correctly chosen so that the radioactivity is minimally
deposited on tubing. Tubing runs to CAMs must be as short as possible.
43
9.21. As CAMs are optimally placed as close as possible to the source of airborne, they are
frequently placed in radiation fields that vary in intensity over time. Therefore, a significant
amount of shielding is necessary to avoid generating incorrect signals due to variations in
local background caused by movement of product, waste, raw materials, etc. Otherwise
packages placed near an insufficiently shielded CAM will appear to cause an increase in
airborne, or mask airborne activities. If the CAM has two detectors, one can be used to correct
the variations in the background. Filter material should be optimally suited to catch the
airbornes. Filter materials include paper and fiberglass for particulate, activated charcoal and
silver zeolite for radioiodine, etc.
Maintenance and calibration
9.22. Following calibration, a label should be attached to the instrument to provide
information, including the organization performing the test, the test certificate number and the
date of the test or the date when the next test is due. Tests should be carried out by an
organization that maintains reference radiation fields traceable to a national standards body.
9.23. Fixed radiation monitoring instruments are not calibrated in the same sense as radiation
survey meters. Since their operation is ‘pass–fail’, fixed instruments should be subject to
periodic operational testing to ensure that they retain the capability to respond to relevant
radiation levels. For example, check sources can be used on a monthly basis to verify that the
radiation room monitor respond appropriately. In some applications, such as in using a single
channel analyzer for air effluent monitoring, the instrument should be calibrated periodically
to ensure that the detector voltage and window settings are still applicable.
9.24. Further information on the establishment and operation of calibration facilities for
radiation survey instruments and recommended calibration procedures is provided in
Ref. [32].
Records of radiation and contamination surveys
9.25. Reports on radiation and contamination levels should include the following information:
Survey date;
Information on the survey instrument (manufacturer, model number and serial
number);
44
Calibration date of the survey instrument;
Correction factors, background subtraction, conversions or other calculations for
the survey instrument if used;
Name of the person performing the survey;
Radiation levels and the corresponding locations are best to record and
communicate on sketches of the section of the building which was surveyed;
Contamination levels and the corresponding locations;
Cause of the contamination, if known;
Any actions taken on the basis of information yielded by the survey.
10. ENVIRONMENTAL MONITORING AND EFFLUENT DISCHARGE
ENVIRONMENTAL MONITORING
10.1. Environmental monitoring of foodstuffs, plants and animals is not a requirement for the
processing facilities covered under this Safety Guide. The environmental monitoring required
is limited to performing and documenting dose rate surveys external to the controlled area,
with the objective to demonstrate that members of the public are receiving effective doses
<1 mSv in a year. In some cases, the boundary to perform these measurements is within the
building. For new facilities, detailed dose rate surveys should be performed, and any
deficiencies in design and construction should be corrected until the facility is deemed safe to
operate under the conditions where maximum dose rates can occur. Once the facility is
operational, routine environmental dose rate surveys should be carried out continuously.
10.2. The environmental verification should be periodically confirmed by measurement of
groundwater or soil samples. Soil samples will always contain trace natural amounts of
radioactivity radioactive isotopes, e.g.
137
Cs, due to atmospheric weapons testing or the
40
naturally occurring K, therefore soil samples should be compared to background soil away
from the discharge stack.
EFFLUENT DISCHARGE
45
10.3. Radioisotope production and processing inherently poses risk of dispersal of radioactive
materials to the environment which can be the primary product or a mixture of decay
products.
10.4. The production technology, the adopted practices and the facility design should all aim
to control the amount of activity routinely discharged and to minimize the risk of discharges.
10.5. Effluent discharges for production facilities should be regulated, based on authorized
discharge limits which should be developed by the operating organization and approved by
the regulatory body. The IAEA publication [33] provides more detailed guidance on
methodology and procedures to develop such authorized discharge limits.
10.6. The effluent streams should be considered carefully prior to planning and construction.
The effluents should also be addressed when planning and implementing new production
lines, when methods or equipment are changed, or when operating conditions of the facility
itself change (ventilation, pressures etc.).
10.7. Effective means should be available for containing releases of activity before they leave
the facility. Best practices include in-process means of capturing and securing gaseous, liquid
and dispersed solid waste. Filtration and trapping systems should be designed to be as close as
possible to the source production in order to minimize the unnecessary contamination of
ducts, piping, etc. The handling of the effluent streams should include safe means of removing
other hazardous components (e.g. air filters might not only be installed to reduce release of
activity into the atmosphere, but also to minimize release of other toxic chemicals).
AIR EFFLUENT MONITORING
10.8. The IAEA publications [33] and [34] have established standards for monitoring air
effluent emissions.
10.9. Quantitative on-line air effluent monitoring should be performed using:
A well shielded detector which views a cross-section of the stack and is oriented not
to detect other sources of radiation;
A well shielded continuous air monitor (CAM) sampling the stack;
A gas flow through chamber detector for inert gases, or other means.
46
However, in all cases a representative sample of the effluent should be taken.
10.10. Off-line measurements should be taken using filters (cartridge or otherwise) which are
replaced daily or weekly (as necessary) and measured.
10.11. If sampling lines are used, the number of bends in tubing should be minimized to
avoid line losses. Tubing material should be correctly chosen so that the deposition of
radioactivity on tubing is minimized.
10.12. Experimental evidence should sometimes be used to validate sampling systems. One
such example is to release an approved activity of
11
C labelled carbon dioxide (11CO2) to
calibrate systems at PET facilities.
10.13. The stability of sampling pump and stack flow rates should be taken into account and
variations may need to be logged.
10.14. Other points should be considered with respect to monitoring of air emissions:
(a)
The emitted activity is product of concentration and air flow.
(b)
The monitor should be capable of measuring relevant radionuclides at sufficient
sensitivity.
(c)
The monitor(s) should be shielded from variations in background radiation.
(d)
If several radionuclides are present, they should, if possible, be identified and
quantified.
LIQUID EFFLUENT MONITORING
10.15. National, regional and municipal regulations should apply to limit the discharge to
liquid effluent streams, in terms of chemical and biological composition, suspended solids,
radioactivity and other hazards.
10.16. Liquid effluent should be monitored on-line or a sample may be taken from a delay
tank. Procedures should be developed to ensure that the delay tank contents are adequately
mixed so that a representative sample may be taken. If an aliquot is to be taken of the sample
(for example for liquid scintillation counting), then the sample also should be agitated to
ensure adequate mixing before the aliquot is taken.
47
MINIMIZING EFFLUENT DISCHARGES
10.17. In planning applications, consideration should be given to the confinement of liquid
borne activity in case of flooding, pipe ruptures or extensive fire-fighting with water.
10.18. The process water should be kept and treated separately. Coolant should only be
diluted with inactive water prior to ultimate disposal. The control of radioactive discharges is
discussed in the IAEA publication [33] including its minimization options.
10.19. Water used for washing and cleaning in isotope production facilities could potentially
be contaminated, depending on the nature of the facility. It might be necessary to pipe such
waste streams to storage tanks, perhaps for decay, but ultimately for analysis, possible
purification/distillation and/or subsequent disposal into the general environment.
10.20. Target and accelerator cooling circuits may become radioactive (excluding the shortlived radionuclide 16N) due to leaching of activated surfaces or from leakages. Therefore, they
should be disposed of only after check of activity.
10.21. Dedicated piping for possibly contaminated/radioactive waste water should be in
place. In case acceptable low limits can be assured under all operating conditions, direct
piping to main sewer can be recommended.
10.22. Workers maintaining such draining installations should be properly protected and
instructed.
10.23. All airstreams in the facility that might contain activity should be considered. This
could include all of the controlled area as well as storage areas, target loading/unloading areas
and potentially also radioisotope generation equipment.
AIR EMISSIONS
10.24. Air monitoring filters should be suitably placed in the ventilation system prior to
exiting the building. If the filter (e.g. charcoal) is in an uncontrolled area, it should be
adequately shielded to minimise the risk of exposure to personnel.
10.25. Corrosive substances (e.g. acids) should not be ventilated through the air monitoring
control system.
48
10.26. The filters in the filtration system should be changed on a regular scheduled basis (e.g.
annually). The frequency of change may need to be increased if an elevated trend in emissions
is observed.
10.27. If radioactive materials are produced that cannot be trapped by the air filtration
system, abatement systems (e.g. exhaust bags) should be utilised to store radioactive materials
until they have decayed to background levels.
10.28. General principles of placement, height of stack, assured ejection speeds and
meteorological considerations should take into account occupied areas and worst case
scenarios, the need for calculation of worst case committed radiation dose to most exposed
member of the public, and the reference to suitable guidelines for this, and possible general
dose constraints (1/10 of annual dose limit to members of the public). Compliance with this is
the responsibility of the facility operator, and could be part of an operation permit.
10.29. Channels, filters and other components should not be attacked by components of the
air stream, nor yield unnecessary particle burdens by themselves (stainless steel or epoxy).
Description of the abundant use of boiling with strong mineral acids should be included as
well
as
good
practices
to
minimize
corrosion
risks
from
the
acid
fumes
(gaswashers/scrubbers).
10.30. Filters bound to contain large activities at any point in time should be placed in
controlled areas and, if appropriate, also shielded or separated from areas of any occupancy.
10.31. Pressure drops and integrity of critical filters should be kept under control by suitable
measures.
10.32. Filters should be removable under radiologically safe conditions (bagging provisions).
10.33. Practices for removal of non-filterable contaminants may include:
(i)
Placement of filters as close as possible to the source, at points of lowest airflow.
(ii)
Use of activated charcoal filters.
(iii)
Use of acid filters/scrubbers.
(iv)
Pressure drops and integrity of critical filters should be kept under control by suitable
measures.
49
(v)
Filters should be removable under radiologically safe conditions (bagging provisions).
(vi)
Ways of testing the efficacy of such filters.
10.34. Considerations concerning non-filterable, non-condensable airborne contaminants
should include:
(i)
Radioactive noble gases.
(ii)
Some PET cyclotron products (examples [11C]-CH4/CO2, [18F]-FCH3 or F2, [13N]-NH3)
cannot be removed from air stream.
(iii)
Tritium, and some tritiated and 14C labelled compounds.
10.35. In case such contaminants pose any significant dose risk/contribution to either workers
or members of the public, measures should be taken to limit and control the release of such
contaminants.
10.36. The most efficient ways is to control the release of contaminants are to contain and
trap the contaminants at the source itself with using gas bags or traps (liquid nitrogen or
cartridges). Another possibility could be tank storage for decay (in case of the PET gases).
11. PERSONAL PROTECTIVE EQUIPMENT
11.1. The operating organization should ensure that engineering controls are in place to
protect workers from exposure to radioisotopes and other associated hazards. In some cases,
even when optimized engineering controls have been implemented, additional protective
measures such as personal protective equipment (PPE) should be used to keep radiation doses
as low as reasonably achievable or to mitigate the consequences of an accident.
11.2. The operating organisation should ensure that workers are provided with suitable and
adequate PPE which meets relevant standards and specifications. According to GSR Part 3
[3], the operating organization is required to provide PPE to workers. The PPE for routine
operations may include:
(a)
Protective clothing, including gloves, overalls and caps for contamination hazards;
50
(b)
Protective respiratory equipment suitable to protect the respiratory tract from the
contamination hazards;
(c)
Protective aprons and gloves and organ shields for external radiation hazards;
(d)
Safety glasses or face shields for splash protection involving radiological liquids.
11.3. The PPE for emergency operations may include:
(a)
Full air suits with air lines or breathing apparatus for entering contaminated areas,
(b)
Lead aprons, critical organ protectors and gloves for handling situations with high
radiation.
11.4. Where appropriate, workers should receive adequate training and refresher training in
the use of PPE. All PPE should be maintained in working order, tested at regular intervals if
appropriate and be maintained for use in the event of usage.
11.5. The reliance on PPE for protection and safety should be minimised by the operating
organisation during normal operations by providing appropriate protective measures and
safety provisions, including well engineered controls and satisfactory working conditions.
11.6. The safety assessment should provide information for the job specification for each area
and process, and during the medical examination it must be determined if a person is
medically fit to use the prescribed PPE for the job. Some of the aspects to be covered for
medical examinations are the possibility of claustrophobia and hypertension, for example, that
would limit the use of some of the PPE.
11.7. Contaminated re-usable PPE like expensive apparels and washed overalls, should be
decayed, and if necessary, decontaminated in a decontamination room. Highly contaminated
PPE should be left to decay before sending for washing. In cases where long-lived
radionuclides are present, the RPO should decide if it can be considered as radioactive waste.
11.8. If the use of PPE is being considered for a task, any additional exposure that could
result owing to the additional time or inconvenience, and any additional non-radiological
risks, should be assessed against performing the task without using PPE.
51
12. NUCLEAR SECURITY CONSIDERATIONS
12.1. The security policy of the organization should aim to deter, detect, delay, and respond to
any attempt to gain or actual unauthorized access to radioactive sources. The following
paragraphs are intended to raise awareness about the security issues that need to be addressed
and which are covered in detail in the IAEA Nuclear Security Series (NSS) publications. In
particular, NSS No. 14 [5] provides recommendations to States and competent authorities on
how to develop or enhance, to implement, and to maintain a nuclear security regime for
radioactive material, associated facilities, and associated activities. NSS No. 11 [6] contains
more specific guidance to assist States in the development of regulatory requirements for the
security of radioactive sources. NSS No. 9 [7] provides guidance on the security of
radioactive material during transportation.
12.2. Nuclear safety and security measures have the common aim of protecting human life,
health and the environment. Safety measures and security measures should be designed and
implemented in a coordinated manner so that security measures do not compromise safety and
safety measures do not compromise security.
12.3. To ensure that safety and security are implemented in a compatible manner, the
government may have designated a responsible body for managing the interfaces between
safety and security in relation to radioactive sources. This may be the regulatory body if the
regulatory body has responsibility for both the safety and security of radioactive sources
under the regulatory infrastructure.
12.4. In radioisotope production, there may be an interface between security and safety
measures with regard to access to information. For safety purposes, information on the
locations and characteristics of radioactive sources and the safety measures in place may need
to be readily accessible. However, this information may also be of potential value to an
adversary, and therefore security considerations may require that the confidentiality of some
sensitive information be protected. Guidance on the protection and confidentiality of sensitive
information in nuclear security is provided in [8]. An appropriate balance needs to be
maintained between the availability of information for safety reasons and the need to protect
sensitive information for security reasons.
12.5. Safety measures designed to prevent the loss of radioactive sources or for protection
against radiation incidents can also provide some benefit against the theft of those sources.
52
For Category 1 sources, for example, it is recommended that measures described in
GSR Part 3 [3] are used. However, the element of intent involved in unauthorized access
means that additional considerations apply for higher activity sources, and additional and/or
different security measures may be needed to protect against unauthorized access.
12.6. The IAEA Nuclear Security Series (NSS) provide guidance on how to define the
requirements for the security of radioactive sources using a graded approach, based on
considerations of threat, the nature of the sources, and the relative attractiveness of the
material for use in a malicious act. NSS No. 11 [6] suggests using the IAEA’s categorization
system in order to assign a particular security level to sources and to help define the necessary
security measures. Radioisotope production sources are typically assigned to Security Level
C, and not higher than Security Level B. The security measures required for each security
function for Security Levels B and C are described in detail in [6].
12.7. It should be noted that, due to their small size, portability and the fact they are most
often used far from any secure facility, radioisotope sources may need additional security
measures or procedures to ensure they remain adequately protected and under control both
during use, transport incidental to use, and while they are not in use. The specific details of
such additional measures will depend on the threat assessment. Reference [6] also contains
illustrative security measures including those for mobile operation where measures applicable
to a fixed installation are not practicable, which can be adapted for mobile security level C
operations.
13. TESTING AND MAINTENANCE OF EQUIPMENT
13.1. To ensure the continued safe operation of the radiation production facility, the operating
organization should set up a formal programme of maintenance and testing to test all safety
functions regularly. The following actions should be performed periodically (or as otherwise
specified below):
(a)
Particular attention should be paid to regular testing of components of the safety
interlock system for correct operation, in accordance with the instructions of the
equipment manufacturer. These tests should be carried out by appropriately qualified
persons in the presence of a RPO.
53
(b)
Periodic leak tests of radioactive sources should be carried out in a manner and at a
frequency as recommended by the source supplier and in accordance with regulatory
requirements.
PERIODIC TESTS
13.2. The ventilation system (buildings, hot cells, fume hoods) should be maintained on a
regular basis (annually).
13.3. The heating/cooling systems, generators, radiation monitoring equipment, interlocks,
freezers, building monitoring system, HEPA filters in clean rooms and dose calibrators should
be maintained on a regular basis. All equipment used in measuring radiation levels, weights
and as required by regulatory agencies should be tested/calibrated/maintained on a regular
basis.
13.4. The following additional tests should be carried out on a monthly basis:
(a)
Check, in accordance with the manufacturer’s instructions, that access to the radiation
room is prevented when the radiation room monitor alarm sounds. Check the emergency
exit procedure by ensuring that the personnel access door can be opened from the inside
and that other means of exit in an emergency are operating properly.
(b)
Check all visual warning signals and alarms for correct operation. Check all control
indicator lights to ensure that they illuminate.
(c)
Verify that all uninteruptible power supplies (UPS)8 are functioning within
specification. It is good to practice a UPS on the cyclotron control system as power
‘dips’ can affect the operation of control units.
(d)
Verify the proper operation of the heat detectors and smoke detectors.
(e)
Verify all safety interlocks on removable shield plugs (or self shield) in the cyclclotron
room.
(f)
Verify that posted notices are in place and that all the details are correct.
8
An uninterruptible power supply is a backup power supply that, in the event of power failure or power
fluctuations, allows enough time for an orderly shutdown of the system or for a standby generator(s) to start up.
54
13.5. If any of the checks indicate a fault or that a safety interlock is not functioning properly,
the facility should not be operated until the system has been returned to its validated
operational state. The return of the facility to normal operation should be subject to approval
by a RPO.
RECORDS
13.6. The results of all tests described above should be recorded on a formal checklist signed
by a RPO who has witnessed the tests.
13.7. The maintenance records should be kept for such periods of time as are prescribed by
the regulatory body.
FACILITY MAINTENANCE AND MODIFICATION
13.8. Maintenance operations at the facility should be coordinated with the manufacturer of
the various pieces of equipment in the facility to ensure that appropriate repairs, modifications
and system upgrades are completed as per approved protocols.
13.9. Bypassing or disabling a safety interlock should be done only with the express, written
approval of a RPO. All circumstances necessitating any component of a safety interlock to be
bypassed or disabled should be documented with a description of the circumstances and the
actions taken, and with the specific approval of a RPO.
13.10. If it becomes necessary to bypass or disable a safety interlock, independent
verification should be obtained either that the accelerator is not on (e.g. ion source is not on).
The affected component of the safety interlock system should only be bypassed or disabled
only long enough to allow entry to the radiation room to remedy the problem (e.g. to repair or
replace the monitor), during which time the relevant portion of the facility will not be in
operation. Entry to the radiation room should be permitted once a satisfactory survey of the
area has been completed.
13.11. If it is necessary to bypass or disable a component of a safety system, the affected
component should be tested for specified operation upon being reinstated. The specific test
will depend on which component is to be tested, but the test should be a duplicate of the
routine test performed to verify specified operation. After verifying that the safety interlocks
55
have been restored to their design function, approval of a RPO should be obtained for a return
to normal operations.
13.12. Since bypassing or disabling any component of the safety interlock system is to be
avoided, except under abnormal circumstances, routine and preventive maintenance functions
should be designed to obviate the necessity for bypassing safety interlocks.
14. RADIOACTIVE WASTE MANAGEMENT AND DECOMMISSIONING
14.1. The Regulatory Body should establish requirements and criteria for radioactive waste.
Radioactive waste is radioactive material for which no further use is foreseen and with
characteristics that make it unsuitable for recycling or authorized discharge. This may include
unsealed and sealed sources [33, 35]. Radioactive waste should be examined in the safety
assessment prior to its generation and needs to have considered non-radiological hazards
(biohazards, chemical content) and the acceptance criteria of the ultimate waste destination
(e.g. national waste site, interim storage site).
14.2. Radioactive waste is generated at certain points in a radioisotope production facility.
Low-level waste is created from contamination control procedures (disposable PPE, cloths,
package, surface and floor swipes, etc.). The waste with the highest concentration is generated
from activated materials within the cyclotron, targets, synthesis processes and quality control
testing. Archive samples and unsold product are other examples of waste.
14.3. The waste management protocols and clearance of materials after processing, storage,
reuse and recycling of material can be effective in reducing the amount of radioactive waste
that requires disposal. The operator has to ensure that these processes are in compliance with
the conditions and criteria established in regulations or by the regulatory body. The regulatory
body also has to ensure that the operator gives due consideration to non-radiological hazards
in applying such options [34].
14.4. The control measures are generally applied in the following order: reduce waste
generation, reuse items as originally intended, recycle materials and, finally, consider disposal
as radioactive waste.
56
14.5. In line with Requirement 10 of GSR Part 6 [36], the production facility is required to
prepare a decommissioning plan (‘from the cradle to the grave’) for their facility which
considers ultimate disposal of all resultant waste, contaminated and/or activated equipment
and materials. This decommissioning plan is required to be periodically reviewed and updated
as necessary in the light of operational experience gained, new or revised safety requirements,
available lessons learned from the decommissioning of similar facilities, and technological
developments relevant to decommissioning [35].
14.6. The production facility may possess sealed sources that will, in time, become spent or
disused sealed sources. They then need to have an approved disposal pathway so that sealed
sources do not become orphaned. The accounting of sealed sources must follow the
requirements of the regulatory body.
14.7. Some production facilities fabricate sealed sources and the radioactive material is
typically in one of three states: raw material, finished product (inventory) or waste. These
production facilities should offer their customers a disposal pathway as a pre-sale condition.
The production facility is responsible for accounting for their sealed sources, and returned,
spent customer sources to document that the sealed sources have not been orphaned.
CHARACTERIZATION OF RADIOACTIVE WASTE
14.8. At radioisotope production facilities, aqueous waste results from chemical processing,
mainly the etching and dissolving of target materials. The waste should only be processed
after its precise characterization. In addition to its radiological, physical, mechanical,
chemical and biological properties, radionuclide impurities from the production process
should be characterized and segregated. Radionuclide impurities in the waste streams shall
first be estimated from predictive models and then measured. Radioactive materials that are
produced in cyclotrons can contain small quantities of longer lived radioisotope impurities
other than the finished product. The production facility is responsible for developing the
waste acceptance criteria for approval by the regulatory body.
14.9. The production facility should follow the clearance criteria established by the regulatory
body. Clearance levels establish at which point material under regulatory control can be
removed from this control [37]. In order to demonstrate that the material in their possession
contains radioactivity below the clearance level, the production facility should first establish
the radioisotopes in the waste streams, and then compare their activity concentrations with the
57
clearance levels. The activity concentrations in waste streams can be determined by
understanding the initial concentrations and calculating for decay and/or by directly
measuring and identifying the activities of the radionuclides present. The production facility
should document this evaluation.
PRINCIPLES OF WASTE MINIMIZATION
14.10. Waste minimization is an important step in waste management and controlling
potential risk as well as cost. The principles of ‘delay and decay’, and ‘concentrate and
contain’ [38] are important in waste minimization.
14.11. Segregation is an important concept in waste minimization within the controlled area.
Waste should be first segregated into two categories: waste that is known or is suspected of
being radioactive, and waste that is believed to be non-radioactive. The latter category should
be verified to meet the clearance criteria.
14.12. Another form of segregation is for biological waste that needs to be treated (by either
autoclaving, sterilizing or incinerating) or pH adjusted of liquids to render them safe (e.g.
iodines must be alkaline).
HANDLING AND PROCESSING OF RADIOACTIVE WASTE
14.13. Depending on local regulatory approval, it may be acceptable to ‘dilute and disperse’
[39] radioactive material. An example of ‘dilute and disperse’ might involve filtered
ventilation exhaust where the activity concentrations of gaseous effluents concentrations that
have been pre-determined (by regulatory approval) not to endanger people or the
environment.
14.14. The facility operator should ensure that radioactive materials and sources from
authorized practices are not discharged to the environment unless:
(a)
Such discharge is within the limits specified in the licence and is carried out in a
controlled manner according to the regulation in force and the authorization issued by
the regulatory body methods; or
(b)
The activity discharged is confirmed to be below clearance or other disposal levels
established by the regulatory body.
14.15. Control measures for release of radioactive materials may include:
58
(a)
Sampling of each batch of waste prior to removal from control.
(b)
If, according to the national policy and strategy, radioactive waste is to be stored in a
centralized storage facility, the operator should adopt provisions to ensure the prompt
transfer of above waste and disused sources to that facility.
OTHER HANDLING GUIDELINES
•
Radioactive waste is characterized in terms of its physical, mechanical, chemical,
radiological and biological properties.
•
Containers for solid wastes should be lined with a durable plastic bag that can be sealed
(tied with plastic adhesive tape or heat-sealed with a radio-frequency welder).
•
If drums of waste are to be compacted at the production facility, the compactor shall be
enclosed to prevent the spread of contamination. Compactor safety must be evaluated to
avoid pinch points, compacting material which could damage the drum, etc.
•
Sharps should be collected separately and stored in rigid, puncture-resistant containers
(preferably metal) that have been clearly labelled ‘sharps’.
•
Refuse cans with lids should be lifted by foot pedals to minimize contamination.
•
Liquids can require chemical adjustment (pH important for radioiodines must remain
alkaline) and immobilization prior to transport.
•
Special precaution may be required on used target foils, target blanks, target bodies and
collimators. The area where target reconditioning is performed needs to shield the
operator’s body and extremities.
ON-SITE STORAGE OF RADIOACTIVE WASTE
14.16. In most production facilities, it is necessary to have a dedicated waste and
contaminated equipment storage room. Access to this room should be secure and ventilated.
Some production facilities place sealed waste containers in air sampling boxes to ensure there
is no volatile radioactivity present prior to disposal.
14.17. Routine contamination and dose rate survey should be performed in the storage room.
An alarming continuous air monitor and respiratory protection may also be used to optimize
safety in this room.
59
14.18. Waste storage locations should be planned and designed to minimize handling,
transport and potential doses to members of the public (if the store room is external to the
building).
PREPARATION OF WASTE SHIPMENTS
14.19. Radioactive waste should be prepared by the production facility to ensure that it is in a
safe and passive form (with regard to radiological, physical, chemical and biological hazards)
before it is placed in an approved transport container to be transferred to the centralized waste
storage facility. This facility should be consulted to determine what type of package, package
contents and configurations are acceptable to be received by them. If the production facility
desires to design, build and test [17] a new waste container, such container has to be
compatible with the handling capabilities of the centralized waste storage facility.
14.20. All floor drains and sinks should discharge into delay/holding tanks and water quality
monitored prior to disposal. The discharge port of the main floor drains should have a
removable bladder type plug to contain the spilled liquid in the drain pipes until it has been
assessed for disposal.
14.21. The contents of section 16 equally apply to waste shipments and should be observed.
15. TRANSPORT OF RADIOACTIVE MATERIAL
TRANSPORT REQUIREMENTS
15.1. Transport of radioactive materials should conform to national regulations inside the
State and IAEA regulations for international transport [20].
Movement within the worksite
15.2. When radioactive materials and sources are to be moved within a site for production
operations, they should be kept in the storage facility until they are to be moved to the new
location.
15.3. The sources should be moved only in shielded containers, and these should be locked
and the keys removed. If a vehicle or trolley is used to move the container, it should be
securely fastened inside the separate compartment of the vehicle. The shielded container
60
should be kept under surveillance for the duration of the movement on the worksite. The keys
of the container should be kept by the authorized person.
Transport to another site
15.4. When radioactive materials are to be transported from the production facility, they
should be kept in the storage facility until they are to be moved to the new site.
15.5. The sources should be moved only in shielded containers, and these should be locked
and the keys removed. The operating organizations should ensure that the transport and the
transport packages comply with the IAEA Regulations for the Safe Transport of Radioactive
Material [20] or equivalent national or international regulations.
15.6. Where applicable, consideration should also be given to binding international
instruments for specific modes of transport, such as by air [40] and by sea [41].
15.7. Regional agreements such as the European Agreement Concerning the International
Carriage of Dangerous Goods by Road (ADR) [42], the European Agreement Concerning the
International Carriage of Dangerous Goods by Inland Waterways (ADN) [43] and the
Agreement of Partial Reach to Facilitate the Transport of Dangerous Goods, Signed by the
Governments of Argentina, Brazil, Paraguay and Uruguay (MERCOSUR/MERCOSUL) [44]
may also apply.
15.8. The IAEA Transport Regulations [20] assign responsibilities for individuals involved in
the transport of radioactive material: the consignor (a person, organization or government that
prepares a consignment for transport), the carrier (the person, organization or government that
undertakes transport of radioactive material) and the consignee (the person, organization or
government that receives a consignment). In some cases, for an operating radioisotope
production facility, the operating organization will perform all three functions and is required
to discharge the responsibilities associated with each function.
15.9. Transport of radioactive material is a complex activity, and a comprehensive overview
of the relevant requirements is outside the scope of this Safety Guide. Guidance on how to
meet transport related requirements is provided in [45].
15.10. Comprehensive recommendations on nuclear security in the transport of radioactive
material are provided in [7].
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16. EMERGENCY PREPAREDNESS AND RESPONSE
GENERAL
16.1. According to GSR Part 3 [3] and GSR Part 7 [13], an emergency is a non-routine
situation that necessitates prompt action, primarily to mitigate a hazard or adverse
consequences for human health and safety, quality of life, property or the environment. This
includes nuclear or radiological emergencies and conventional emergencies such as fires,
release of hazardous chemicals, storms or earthquakes. It includes situations for which prompt
action is warranted to mitigate the adverse effects of a perceived hazard.
16.2. As defined in GSR Part 3 [3] and GSR Part 7 [13], a nuclear or radiological emergency
is an emergency in which there is, or is perceived to be, a hazard due to:
(a)
The energy resulting from a nuclear chain reaction or from the decay of the products of
a chain reaction; or
(b)
Radiation exposure.
16.3. Incidents and accidents at a radioisotope production can occur mainly as a result of
operator error or equipment failure and may lead to a radiological emergency. Typical
incidents and accidents include: (1) target package breach; (2) higher dose rate than expected;
(3) dropped source; (4) leaking source; (5) fire inside the hot cell/clean rooms/other
production areas; (6) loss of supply air to the facility and/or loss of exhaust air from the hot
cells; (7) breakage of the cooling line for the cyclotron and the targetary system and
consequent flooding in the facility; (8) natural disasters (e.g. hurricane) affecting the facility;
and (9) nuclear security events resulting in loss of control over radioactive material or the
facility, such as theft or sabotage of radioactive material.
16.4. The hazards associated with the operation of a radioisotope production facility and the
potential consequences of a radiological emergency are required to be assessed as a means to
provide a basis for establishing adequate arrangements for emergency preparedness and
response [13, 44]. Potential emergencies that could affect workers, members of the public or
the environment and could warrant emergency response actions should be identified in the
operating organization’s hazard assessment [13, 46].
16.5. Based on the assessed hazards and potential consequences, emergency arrangements
should be established for the radioisotope production facility in accordance with Refs [47–
62
49]. Radioisotope production facilities generally fall into emergency preparedness category III
described in [13, 47]. Emergency arrangements that correspond to this category should be
established for preparedness and response for a radiological emergency involving the
radioisotope production facility. Some radioisotope production facilities may pose limited
hazards on-site and off-site. However, addressing the perceived hazards or other nonradiological hazards in these circumstances may warrant implementing parts of emergency
arrangements.
16.6. The applicability of various sections of GSR Part 7 to Emergency Preparedness
Category III is listed in the Table in Annex 1 to GSR Part 7 and these should be used during
the preparation of EPR plans for the facility.
EMERGENCY PLANS AND PROCEDURES
16.7. Although prevention of incidents and accidents is the first line of defence, emergencies
still may occur. Operating organizations are required to have in place an emergency plan and
procedures developed at the preparedness stage [13], so as to be able to respond effectively to
an emergency involving the facility under their responsibility.
16.8. An outline for facility (on-site) emergency plan that should be used for developing an
emergency plan of the radioisotope production facility can be found in Ref. [48]. Notices
outlining the notification and activation procedures in case of an emergency may be clearly
and visibly posted inside the facility at locations where they might be needed, and staff should
be trained in these procedures (see para. 4.2.18 of Ref. [48]).
16.9. The emergency plan for a radioisotope production facility should include, but not be
limited to, scenarios such as theft of sources, on-site contamination or leaking due to damage
of the source, accidental releases into the environment and overexposures of workers.
Emergency procedures should include, but not be limited to: (a) notification and activation
protocols; (b) communication and coordination arrangements; (c) provisions for obtaining
support from off-site emergency services; (d) provision of instructions to the site personnel
and provisions for accounting the site personnel; (e) delineation of the affected area and
access control; (f) measures and actions to protect site personnel and emergency workers; and
(g) arrangements for communication with the public etc. A qualified expert/RPA may be
consulted, where possible, when drawing up emergency plans and procedures. Examples of
63
immediate actions to be taken in case of a radiological emergency at a radioisotope
production facility are given in Annex II.
16.10. Requirements and recommendations for on-site and off-site emergency preparedness
and response are given in the IAEA safety standards [13, 46, 47] and Schedule IV of Ref. [3].
Technical guidance on developing adequate emergency arrangements at the organizational,
local and national levels on a step by step basis is also available from the IAEA [46].
Technical guidance regarding generic procedures for assessment and response during a
radiological emergency is also available [49].
16.11. Implementation of the on-site emergency plan and procedures may require off-site
support (e.g. off-site response organizations, emergency services, radiation protection
specialists) as addressed in Refs [13, 46]. The emergency plan should elaborate arrangements
for obtaining such off-site support.
16.12. Operating organizations are required to submit for approval their on-site emergency
plans to the regulatory body [13]. This is to be done when applying for an authorization.
16.13. Emergency plans and procedures are required to be periodically reviewed and updated
with the aim to incorporate lessons from research, operating experience (such as response to
emergencies) and from exercises [13].
EMERGENCY EQUIPMENT
16.14. Operators are required to ensure that all necessary tools, supplies, equipment,
communication systems, facilities and documentation for responding to emergencies is made
available and are subjected to a quality management programme which includes inventories,
resupplies, tests and calibrations [13]. All necessary tools, supplies, equipment,
communication systems, facilities and documentation should be maintained in a manner that
is readily available and functional for use under emergency conditions.
16.15. For emergencies involving radioisotope production sources, the following equipment
should be considered, as appropriate:
Appropriate and functional survey meters to measure both high and low dose rates;
Personal alarm dosimeters and direct reading dosimeters (preferably electronic
personal dosimeters);
64
Additional personal dosimeters (OSL dosimeters, thermoluminescent dosimeters
and/or film badges);
Barrier materials and notices;
Lead bricks;
Suitable tool kit and source recovery equipment (long handling tongs, pliers,
screwdrivers, bolt cutters, adjustable spanner, torch, lead source storage container);
Materials and agents for decontamination [50];
Spare shielded container;
Plastic sheets, air tight bags for rupture of gaseous sources, swipe test kit, measuring
tape;
Communication equipment (e.g. mobile phones);
Spare batteries for survey meters, electronic personal dosimeters, mobile phones and
torches;
Pens, paper, calculator and an incident log book with first responder sheets;
Equipment manuals, procedures, instructions.
16.16. If it is suspected that a radioactive source might have been damaged, consideration
should be given to detect the leak promptly and to assess the contamination before being
further spread out.
TRAINING AND EXERCISES
16.17. All personnel who have role and responsibilities in an emergency response are
required to be designated emergency workers and to be adequately qualified and trained for
the effective fulfilment of their duties [13]. This should include both familiarization with and
understanding of the plans, procedures, analytical tools and other arrangements, together with
specific training on implementing specific emergency procedures and on the use of the
emergency equipment, as appropriate. This is required to include guidance and training on the
approximate radius of the inner cordoned off area in which urgent protective actions would
initially be taken, and on the adjustment of this area on the basis of observed or assessed
65
conditions on the site [13]. Provisions for training should be reviewed periodically to ensure
the continued proficiency of emergency workers.
16.18. Designated emergency workers should implement only those parts of the emergency
plans or those emergency procedures for which they have been given authority and
responsibility and for which they have been trained.
16.19. Exercise programmes are required to be developed and implemented to ensure that all
specified functions in an emergency response as well as organizational interfaces are tested at
suitable intervals [13]. Technical guidance on preparation, conduct and evaluation of
exercises including technical guidance on various types of exercises, their purpose as well as
examples of scenarios for category III facilities can be found in Ref. [47].
16.20. Staff should be trained appropriately in emergency response, including:
(i)
Recognizing the circumstances indicative of an emergency situation;
(ii)
Notification and activation procedures including provisions for obtaining
assistance from off-site emergency services;
(iii)
Implementation of necessary on-site mitigatory actions and protective actions onsite including provision of first aid and evacuation procedures for non-essential
personnel;
(iv)
Assessment of the situation;
(v)
Use of emergency response tools and equipment including fire extinguishing gear
and the rules of engagement;
(vi)
Implementation of recovery actions including decontamination;
(vii) Measures to be followed for their protection during the emergency response.
16.21. Any lessons learned are required to be fed back into reviews and, as necessary,
revisions of the emergency plans and procedures [13].
REPORTING
16.22. Arrangements are required to be made to undertake a timely and comprehensive
analysis of an emergency and the emergency response [13]. A comprehensive report on the
66
findings of the analysis should be prepared by the RPO in consultation, as appropriate, with
relevant interested parties and, if necessary, with qualified expert(s)/RPA(s).
16.23. The report should be submitted to senior management as well as to the regulatory body
and, as appropriate, to other relevant authorities at local, regional or national level. If the
emergency could have been caused by an equipment malfunction, the supplier and other users
of similar equipment should be promptly informed so that the equipment can be evaluated and
appropriate action taken and similar emergencies avoided.
16.24. The report should, inter alia, include:
(a)
A detailed description of the emergency including specifics of the equipment and
sources involved;
(b)
Environmental and working conditions at the time of the emergency, with particular
reference to whether or not these conditions played any significant part in causing the
emergency or affecting the outcome;
(c)
The root causes of the emergency;
(d)
A detailed description of the emergency response taken;
(e)
Personnel involved, the work they carried out, their skills and qualifications;
(f)
An assessment and summary of the doses received by all affected individuals;
(g)
Corrective actions identified with the aim of preventing similar emergencies in the
future and necessary for improving overall radiation safety, security and emergency
arrangements; and
(h)
Proposed means and timeframes for implementation of the corrective actions identified
and responsible staff.
67
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HEALTH ORGANIZATION, UNITED NATIONS ENVIROMENT PROGRAMME, WORLD
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Material, IAEA Nuclear Security Series No. 9, IAEA, Vienna (2008).
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Nuclear Security Series No. 23-G, IAEA, Vienna (2015).
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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).
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IAEA Safety Standards Series No. RS-G-1.9, IAEA, Vienna (2005).
INTERNATIONAL ATOMIC ENERGY AGENCY, Dangerous Quantities of Radioactive
Material (D-Values), IAEA, Vienna (2006).
INTERNATIONAL ATOMIC ENERGY AGENCY, Cyclotron Produced Radionuclides:
Guidelines for Setting Up a Facility, Technical Reports Series No. 471, IAEA, Vienna (2009).
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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, UNITED NATIONS ENVIRONMENT PROGRAMME,
UNITED NATIONS OFFICE FOR THE CO-ORDINATION OF HUMANITARIAN
AFFAIRS, WORLD HEALTH ORGANIZATION, WORLD METEOROLOGICAL
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70
ANNEX I
KEY RADIATION PROTECTION ISSUES TO BE TAKEN INTO ACCOUNT WHEN
PLANNING THE PRODUCTION OF CYCLOTRON RADIOISOTOPES
At the outset, the accelerator building design should comply with radiation safety requirements on
protection of workers and public. Some of the key requirements are listed below:
1. Material, process, and personnel flow diagrams are important for the design of the facility
2. Appropriate shielding
3. Carefully designed mechanical, electrical, and utility requirements for the operation of the
cyclotron in the vault
4. Negative pressure in the cyclotron vault
5. Adequately shielded hot cells
6. Air handling requirements for the facility
7. Air pressure regimes in rooms and hot cells
8. Radiation monitoring
9. Automated response system for engineering controls in the building
10. Security of radioactive materials
11. Decommissioning plan
12. Health and safety requirements (fire protection, etc.)
13. Utility capacity (e.g. electric power, coolant, medical gases etc.)
14. R&D requirements
15. GMP requirements
16. Receipt quarantine of materials
17. Emergency planning and response
18. IT capacities and networking
19. Redundancy
20. Quality control laboratories
71
ANNEX II
EXAMPLE OF IMMEDIATE ON-SITE RESPONSE ACTIONS IN CASE OF AN
EMERGENCY AT A RADIOISOTOPE PRODUCTION FACILITY
II.I.
This Annex provides practical guidance for immediate on-site response actions that
may be warranted to be taken in case of an emergency at a radioisotope production facility by
operating personnel and/or the RPO. Although the actions are listed in the sequence in which
they can be expected to generally be performed, it may be necessary that they be implemented
in another sequence or simultaneously. These actions are generic and focused only on those
that are immediately warranted on-site. They do not account for all the emergency response
actions that may be warranted off-site and for those actions that may be warranted beyond
these immediate actions on-site as required in EPR related Safety Standards [II-1–II-IV].
Gamma and neutrons
II.II.
(a)
Operating personnel:
Recognizes promptly abnormal conditions at the site that is indicative of an emergency
and activates the pre-planned emergency response;
(b)
Takes lifesaving actions and gives first aid;
(c)
Evacuates non-essential personnel and visitors from the potentially hazardous area;
(d)
Establishes inner cordoned off area and prevents any access;
(e)
Notifies relevant authorities (on-site and off-site) including the radiation protection
officer (RPO);
(f)
Measures the radiation dose rates and records any doses measured by direct reading
dosimeters;
(g)
Re-adjusts the inner cordoned off area;
(h)
Keeps the area always attended until the respective emergency workers arrive.
II.III. The RPO:
(a)
Monitors on-site personnel and visitors for contamination and ensures contaminated
individuals and items do not leave the site undetected;
72
(b)
Recommends decontamination of individuals and items, as appropriate, following
respective emergency procedures;
(c)
Confirms that off-site protective actions are not needed;
(d)
Ensures unified command and control system is established as pre-planned to manage
the emergency response;
(e)
Recommends a specific course of action on the basis of previously established
emergency procedures, taking care to adequately protect emergency workers and on-site
personnel as well as to minimize their doses;
(f)
If needed, rehearses the planned course of action with respective emergency workers
before entering the inner cordoned off area to implement the emergency plan;
(g)
Implements, along with designated emergency workers, the planned course of action;
(h)
If necessary, calls for technical assistance from a qualified expert/RPA and/or from the
manufacturer of equipment;
(i)
Ensures that the access control to the inner cordoned off area is in place at all times;
(j)
As appropriate, notifies senior management and the regulatory body and ensures
continuous communication with off-site authorities.
REFERENCES TO ANNEX II
II-1. INTERNATIONAL ATOMIC ENERGY AGENCY, Arrangements for Preparedness for a
Nuclear or Radiological Emergency, IAEA Safety Standards Series No. GS-G-2.1, IAEA,
Vienna (2007).
II-II. FOOD AND AGRICULTURE ORGANIZATION OF THE UNITED NATIONS,
INTERNATIONAL ATOMIC ENERGY AGENCY, INTERNATIONAL LABOUR OFFICE,
PAN AMERICAN HEALTH ORGANIZATION AND WORLD HEALTH ORGANIZATION,
Criteria for Use in Preparedness and Response for a Nuclear or Radiological Emergency:
General Safety Guide, IAEA Safety Standards Series No. GSG-2, IAEA, Vienna (2011).
II-III. INTERNATIONAL ATOMIC ENERGY AGENCY, Method for Developing Arrangements for
Response to a Nuclear or Radiological Emergency, EPR-Method (2003), IAEA, Vienna (2003).
II-IV. INTERNATIONAL ATOMIC ENERGY AGENCY, Generic Procedures for Assessment and
Response during a Radiological Emergency, IAEA-TECDOC-1162, IAEA, Vienna (2000).
73
CONTRIBUTORS TO DRAFTING AND REVIEW
Blackley, R.
Australian Nuclear Science and Technology Organisation, Australia
Castellanos Macchiorlato, A.
National Atomic Energy Commission, Argentina
Fisher, D.
Pacific Northwest National Laboratory, United States of America
Geets, J-M.
Ion Beam Applications s.a., Belgium
Gusev, I.
International Atomic Energy Agency
Hertgers, K.
The Netherlands
Jensen, M.
Riso National Laboratory for Sustainable Energy, Denmark
Karev, A.
Federal Medical Biophysical Center of A.I. Burnazyana, Russian
Federation
Kochnov, O
Research and Scientific Institute of Physical Chemistry, Russian
Federation
Mukherjee, B.
Duisburg-Essen University, Germany
Nauser, T.
Eidgenössische Technische Hochschule, Switzerland
O'Donnell, R.
Ireland
Rajashekharrao, B.
Bhabha Atomic Research Centre, India
Reber, E.
International Atomic Energy Agency
Utkin, K.
State Atomic Energy "ROSATOM", Russian Federation
CONSULTANTS MEETINGS
2–4 December 2008, IAEA, Vienna
9–13 December 2013, IAEA, Vienna
20–24 April 2015, IAEA, Vienna
28 September–2 October 2015, IAEA, Vienna
74
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| File Created | 2017-04-18 |