DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 3.1

Prepared by Safety and Radiation Science

Australian Nuclear Science and Technology Organisation

With the assistance of consultants in Australia and the United Kingdom, and IAEA experts from Korea, Mongolia, New Zealand, the Philippines and Thailand

Use of this material, acknowledging the IAEA and ANSTO as the source, is permitted for non-profit training

Acknowledgements

This material has been prepared by an extensive team of international experts from the fields of radiation protection, medicine and education.  It is not possible to mention by name all those who have had an input to; the initial planning; writing modules and workbooks; preparing and marking final assignments; proof-reading and editing; using the material and providing valuable feedback; organizing and hosting review meetings; and general administrative support.

The support of ANSTO management in providing the resource and encouragement is gratefully acknowledged.

The enthusiastic support of IAEA Technical Officers has been vital to the success of this project.

PART 3

RADIATION PROTECTION INFRASTRUCTURE

MODULE 3.1

IONIZING RADIATION LEGISLATION AND REGULATIONS

CONTENTS

OVERVIEW

In this module you will learn about the national infrastructure necessary to ensure that acceptable standards of radiation protection are maintained in each country.  As this module builds on the knowledge from Module 2.1, Principles of Radiation Protection, it may be useful for you to revise that module before you begin.

The national infrastructure in your country should include all organizations and individual persons responsible for radiation protection.  However, the national government is responsible for establishing a legal framework to set the radiation protection regulations of the country.  These regulations should be enforced by a regulatory authority to ensure national radiation protection.  For adequate protection of your workplace, you must be aware of your country's legal requirements for radiation protection.  If you are authorized by your country's regulatory authority to perform your work, then you are also responsible for meeting its requirements.  It is therefore important that you understand the national infrastructure for radiation protection in your country and its regulatory requirements.

There are three assessable research assignments in this module in which you will be asked to compare what you have learnt with what exists in your country.  It is therefore important that you have information about your country’s radiation protection legislation and regulations available as you work through this module.

 

It is estimated that you will need to spend eight to ten hours to complete this module before commencing the assignment.  The estimated time is for your guidance only.  Do not worry if you need to spend more time to understand the module.

MATERIALS

In addition to your writing materials and workbook, you will need a copy of your country's radiation protection legislation and regulations while studying this module.

NOW YOU ARE READY TO START WORK

 1.NATIONAL INFRASTRUCTURE

The national infrastructure for radiation protection should provide the basic organization and management needed to ensure adequate protection and safety of occupationally exposed workers and members of the public.  One of its main responsibilities is to support and guide the actions of licensed and registered owners of radioactive materials to ensure that radiation safety is maintained.  In addition, the national infrastructure should also ensure the maintenance of expertise in radiation protection, and provide protection in cases where no other organization is responsible, e.g. where hazards are due to natural sources or past practices.

Due to the wide ranging responsibilities of the national infrastructure, the basic organization should be composed of a number of essential parts each with their own functions and responsibilities.  These parts are discussed in detail in the following sections.

 1.1Parts of a National Infrastructure

Essentially, a national infrastructure should consist of five main parts.  These are shown in Figure 1.

Figure 1

The Parts of a National Infrastructure

1.2Responsibilities of a National Infrastructure

Each part has an important role to play in the overall maintenance of radiation protection of a particular country and the responsibilities of each of these part are as follows:

0. The legal framework should provide the basis for the regulation of practices and should assign the responsibilities for radiation protection.

1. The regulatory authority of a country should enforce these legal requirements.

2. The provision of resources for interventions should ensure suitable response to situations beyond the capabilities or responsibilities of licensees and registrants or other organizations.

3. The provision of education, training and information should maintain the national expertise and inform both organizations and the public on matters relating to radiation protection.

4. The provision of facilities and services beyond the resources of licensees and registrants should ensure that suitable safety standards are met and maintained.

SELF-CHECK 1

Now see how much you have understood by answering the following questions in your work book:

1. What are the reasons for having a national infrastructure for radiation protection?

2. What are the essential parts of a national infrastructure?

3. What are the main functions and responsibilities of each part?

Now check your answers with the model answers in your workbook.

RESEARCH ASSIGNMENT 1

Briefly compare the national infrastructure described in this module with the national infrastructure for radiation protection in your country.  For this assignment, you may need to contact your supervisor or local government department.  Your response should be no longer than 300 words and should be written in your workbook.

 2.REGULATIONS AND THE LEGAL FRAMEWORK

The provision of a legal framework is the responsibility of the government of a country.  The legal framework provides regulations which set out the safety requirements for radiation protection of occupationally exposed workers as well as members of the public.  Essentially, these regulations should set the safety requirements by defining the basis upon which a person or organization may be authorized to use sources of ionizing radiation and the safety criteria which must be met during the course of such use.  They should also assign the responsibilities for meeting these requirements (see Figure 2).

Figure 2

Responsibilities of The Legal Framework

Although national regulations may vary from country to country, these regulations are usually based on the recommendations of the ICRP and the BSS.  As already mentioned in Module 2.1, Principles of Radiation Protection, these recommendations have no international legal standing but they can serve as a suitable basis for radiation protection legislation if incorporated into the regulations of a country.  This is particularly true of the BSS as they are more practical and hence more suitable for adoption as regulations.  The following sections introduce the general safety requirements as described by the BSS.

 2.1Authorization

Regulations provide for the authorization of persons to perform practices involving sources of radiation.  Authorizations are required for the following practices:

• Possession and intended use of any source of radiation including the processing, manufacture, storage, repair, maintenance and transport of radioactive materials.

• Use of any source of radiation including commissioning, testing, demonstrating, and use in its intended function.

• Transfer of any radiation source including sale, loan, hire, donation, disposal and abandonment.

• Any other practice specified by the regulatory authority.

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 2.1.1Notification

Any person, group or organization wishing to commence a practice for which an authorization may be required must notify the regulatory authority of that intention.  Where necessary, an application for the appropriate authorization must be submitted to the regulatory authority.  Where the applicant is already authorized, an application for alteration of the existing authorization may be required.  Once issued, authorizations are valid for a set time period, after which the holder must apply for a renewal of the authorization.  Conditions may be attached by the regulatory authority to the authorization, compliance with which is required by the regulations.

 2.1.2Application, alteration and renewal

Procedures for application for a new authorization, or an alteration to or renewal of an existing authorization, are specified in the regulations.  In addition to submitting an application, the applicant must do the following:

• Submit to the regulatory authority all information required for assessment of the application.  The regulatory authority will advise on what information is required, where necessary.

• Refrain from commencing a new or modified practice which is the subject of an application.

• Assess the extent of hazards associated with an intended new or modified practice, including potential hazards, and arrange for adequate protection and safety of workers and members of the public.

• Where required by the regulatory authority, complete a safety assessment of the practice and submit the assessment as part of the application (see Section 3.1 of this module).

Applicants for authorization of a practice where sources are to be used for medical purposes must also submit the following information:

• The qualifications in radiation protection of the medical practitioners (or equivalent) who are to designated by name in the authorization.

• A statement that only medical practitioners (or equivalent) who are to be designated by name in the authorization or specified in the regulations will be permitted to prescribe medical exposure by the practice.

 2.1.3Licensing and registration

Authorizations may be in the form of licences or registrations.  In general, a licence is required where the safety of the practice is mainly dependent on the person conducting the practice.  In cases where the safety of the practice is ensured by the design of facilities or equipment, operating procedures are simple to follow, training requirements are minimal and few problems are known to be associated with the practice, a registration may be issued.  This concept is illustrated in Figure 3.

Figure 3

Licensing and Registration

Authorizations are issued to a licensee or registrant, who may be an individual, a group or an organization.  If an individual is not named as the authorized person, the licence or registration must specifically identify the group or organization responsible for ensuring compliance with the regulatory requirements.  Licensees and registrants bear the overall responsibility for ensuring the protection of persons and the safety of sources for which they are authorized.  Other people may be appointed to perform tasks related to protection and safety, e.g. radiation protection officers but the licensee or registrant retains the responsibility for these tasks.

 2.1.4Exemption of practices and sources

Not all practices and sources will give rise to radiation hazards requiring regulatory control.  Exemptions of practices (and sources within a practice) from the requirements of the regulations are necessary to ensure that resources are not wasted in regulating and controlling negligible hazards.  Exempt practices and sources are usually specified in the regulations.  The BSS provides general principles for exemption as follows:

• the radiation risk from the exempted practice or source must be of no concern to the regulatory authority;

• the collective radiological impact of the exempted practice or source must be sufficiently low as to not warrant regulatory control; and

• the exempted practice or source must be inherently safe i.e. there is negligible likelihood that an accident could lead to risks of concern to the regulatory authority.

To simplify application of the above principles, the BSS suggest that practices or sources may be considered exempt if they meet the following exemption criteria in all feasible situations:

• the effective dose expected to be incurred by any member of the public due to the exempted practice or source is around 0.01 mSv or less in a year; and

• either the collective effective dose committed by one year of performance of the practice is no more than about 1 person Sv or an assessment for the optimization of protection shows that exemption is the best option.

This is illustrated by Figure 4.

Figure 4

Exemption Criteria for Sources and Practices

As a further guide to the exemption of sources within practices, the BSS provide activities of radioactive substances, and dose rates and energies of radiation produced by radiation generators below which the source may be exempted without further consideration of the principles for exemption.

SELF-CHECK 2

Now see how much you have understood by answering the following questions in your workbook:

1. What are the three general types of practice for which authorizations may be required?

2. What is the difference between a licence and a registration?

3. What are the principles by which a practice or source may be exempt from the requirements of regulations?

Now check your answers with the model answers given in your workbook.

 2.2Responsibilities

Another important aspect of national regulations is to assign the responsibilities for radiation protection to certain people or organizations. Licensees and registrants have a number of general and specific responsibilities, as set out in the regulations.  In some practices, these responsibilities are met by the employer, or the licensee or registrant may appoint a Radiation Protection Officer (RPO) to oversee the application of the regulatory requirements at a particular practice.  Workers at a practice also have responsibilities for the protection and safety of themselves and other workers.

 2.2.1Responsibilities of licensees, registrants and employers

The main responsibilities of licensees, registrants and employers are as follows:

• To establish objectives for radiation protection in compliance with the requirements of the regulations; and

• To develop, implement and document a radiation protection program to ensure compliance with the regulations.

Other general responsibilities, within the radiation protection program, are as follows:  

• To determine the resources and actions needed to achieve the objectives of the program; and to ensure that the resources are provided and the actions are properly performed.

• To regularly review the resources and actions and verify that the radiation protection objectives are being achieved.

• To identify any deficiencies in the resources and actions, to take corrective measures, and to prevent their recurrence.

• To establish arrangements to facilitate the communication and cooperation of all persons involved in radiation protection.

• To keep appropriate records to demonstrate the discharge of the responsibilities.

 2.2.2Responsibilities of radiation protection officers

In some cases, the licensee or registrant may appoint a Radiation Protection Officer (RPO) to oversee the application of the regulatory requirements at a particular practice.  The RPO must be technically competent in radiation protection, and the appointment of the RPO must be approved by the regulatory authority.  The general responsibilities of an RPO are as follows:

• To provide technical advice to the licensee or registrant on radiation protection.

• To facilitatecompliance with the regulations.

• To provide advice and arrange appropriate radiation protection training for other workers.

This is illustrated by Figure 5.

Figure 5

The Responsibilities of a Radiation Protection Officer

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 2.2.3Responsibilities of workers

Workers have responsibilities for the protection and safety of themselves and other workers.  The general responsibilities of workers are as follows:

• To follow any applicable rules and procedures for radiation protection, including the use of monitoring equipment and protective clothing.

• To cooperate with the operation of the radiation protection program.

• To provide to the employer, licensee or registrant any information necessary to ensure effective and comprehensive radiation protection for themselves and others.

• To abstain from any wilful action that could result in contravention of the regulations.

• To accept the training and instruction necessary to ensure compliance with the regulations.

SELF-CHECK 3

Now see how much you have understood by answering the following questions in your workbook:

1. What are the two main responsibilities of licensees and registrants?

2. What are the general responsibilities of a Radiation Protection Officer?

3. Give three examples of general responsibilities of workers at a practice.

Now check your answers with the model answers given in your workbook.

RESEARCH ASSIGNMENT 2

Briefly compare the radiation protection regulations and licensing system in your country to that described in this module.  Include comparisons of the types of authorizations issued, the types of practices which require authorization, whether there are any exemptions to authorization and details of how authorizations are obtained, altered and renewed.  You should also outline the responsibilities of licensees/registrants or employers, Radiation Protection Officers and workers as detailed in the regulations and legislation of your country.  For this assignment, you will need a copy of the radiation protection legislation and regulations of your country and/or state.  Your response should be no longer than 600 words and should be written in your workbook.

 3.REGULATORY AUTHORITY

The legal framework for radiation protection must establish a regulatory authority to ensure compliance with the regulations.  The legal framework must also provide for the separation of the regulatory authority from organizations that conduct or promote activities involving radiation exposure, to ensure that the independence of the regulatory authority is maintained.  The regulatory authority must be provided with sufficient powers and resources to perform its functions, which are as follows:

• To assess applications for authorizations of practices and to issue, renew and alter authorizations as appropriate.

• To inspect authorized practices to confirm that regulatory requirements are being met.

• To enforce the regulatory requirements where non-compliance with the requirements are found.

• To encourage the development of a safety culture.

This is illustrated by Figure 6.

Figure 6

The Responsibilities of a Regulatory Authority

The form and structure of regulatory authorities can vary widely.  In some cases, separate authorities regulate different practices, e.g. different authorities may regulate the use of X-ray equipment and radioactive materials.  The following sections describe the functions of the regulatory authority, assuming that a single authority regulates all practices.

 3.1Assessment

The regulatory authority must perform a safety assessment of all applications for authorizations to establish whether the requirements for radiation protection and safety can be met at the practice.  The extent of the safety assessment can vary according to the hazard associated by the practice.  For practices where hazards are low, e.g. use of baggage X-ray equipment, a generic safety assessment of classes of practice will be sufficient.  At more hazardous practices, e.g. nuclear facilities, separate safety assessments may be required for different stages of the practice, including siting, design, construction, commissioning, operation and decommissioning.  The important aspects involved in performing a safety assessment are discussed later in Module 3.3, Radiological Safety Assessment.

Once a safety assessment has been completed, the regulatory authority must assess the information for compliance with the appropriate regulatory requirements and national and international standards.  The extent of the assessment will depend on the hazard presented by the practice.  Where necessary, the applicant will be requested to provide additional information, and inspections, tests or certifications may be required.

 3.2Inspections

After an authorization has been issued, the regulatory authority must confirm compliance with the regulations by periodic inspections of the practice.  The frequency and detail of inspections should be determined by the extent of the hazard associated with the practice, and sometimes, by the history of compliance of the practice.

The duties and responsibilities of inspectors must be set out in the regulations to enable them to take actions necessary to assess compliance with the regulations.  The inspectors may also have to take immediate action to ensure the safety of workers as well as the public.  The general duties and responsibilities of inspectors are as follows:

• To enter premises where authorized practices or sources are located at a time convenient for the inspector and the licensee, registrant or employer, for the purpose of assessing compliance with regulations.

• To search premises for evidence of non-compliance with regulations.

• To examine records of a practice or source, and make copies of those records where necessary.

• To interview licensees, registrants, employers and workers regarding safety related issues.

• To test equipment, collect samples and analyse materials to assess compliance.

• To order the suspension of any practice, remove any source, or require any action to be undertaken to ensure adequate protection and safety where the inspector considers that there is an unacceptable hazard associated with a practice or source.

The regulations must also provide for the assistance and protection of inspectors in performing their duties.  To ensure that inspections are consistent between different inspectors and different practices, the regulatory authority should establish inspection criteria and procedures.  These should be made available to authorized persons to assist them in ensuring compliance with the regulations and inspection criteria.

 3.3Enforcement

Where an inspection reveals a non-compliance with the regulations and requirements of the regulatory authority, the authority may take enforcement actions to ensure that compliance is resumed and maintained.  The severity of the enforcement depends on the safety significance of the non-compliance and, sometimes, on whether the non-compliance was discovered by an inspector or reported by the licensee, registrant or employer.  Enforcement may include fines or other measures which are designed to encourage the authorized persons to avoid non-compliance with the regulations.  The regulations should also allow the authorized persons to appeal against the imposition or severity of enforcement actions or fines.

The regulations should require the authorized persons to take actions themselves to ensure compliance.  These include the following actions:

• Investigate the cause, circumstances and consequences of the non-compliance.

• Take all actions necessary to correct the non-compliance, and seek to prevent it from happening again.

• Notify the regulatory authority of the cause and consequences of the non-compliance, and corrective and preventative actions taken or planned.

Requirements for the notification of the regulatory authority of non-compliance with regulations should be specified in the regulations.  These requirements should vary with the severity of the non-compliance, e.g. where an emergency exposure has occurred, or is likely to occur, the regulatory authority should be notified promptly.

It is preferable that corrective and preventative actions following a non-compliance result from within the practice, rather than being imposed by the regulatory authority.  For this reason, enforcement actions are often less severe if the authorized person has notified the regulatory authority of a non-compliance, and has taken corrective and preventative actions themselves.

Communication and cooperation between the regulatory authority and the authorized persons should be encouraged.  The regulatory authority should be willing to provide guidance on meeting the regulations, and authorized persons should cooperate in providing information and practices and sources, and access to inspectors.

 3.4Encouragement of Safety Culture

Safety culture is the attitude of individuals and organizations to radiation protection and safety issues.  Where there is a good safety culture, such issues receive appropriate attention without the intervention of the regulatory authority.  The need for a good safety culture has been demonstrated by accidents such as the Chernobyl nuclear power plant accident, where poor safety culture was found to be one of the contributing factors that led to the accident.

The licensee, registrant or employer is responsible for the development of safety culture at a practice.  The regulatory authority cannot impose safety culture, but it should encourage the development of the following safety culture characteristics:

• Commitment of workers, authorized persons and the regulatory authority to safety.

• Accountability of all individuals for protection and safety, including those at senior management level.

• Positive attitudes to safety which include a willingness to question and learn, and to discourage complacency on safety issues.

• Communication on safety issues throughout the organization by the authorized persons.

By developing these characteristics, a good safety culture should be established and maintained.  This should ensure a safe working environment and in turn, avoid accident situations.

SELF-CHECK 4

Now see how much you have understood by answering the following questions in your workbook.

1. What are the functions of the regulatory authority?

2. Why should a safety assessment be performed for each application for authorization?

3. How should a regulatory authority confirm compliance with the regulations?

4. Give three examples of the duties and responsibilities of an inspector.

5. When may enforcement of the regulations be necessary?

6. Why is it important that a good safety culture is established and maintained within a practice?

Now check your answers with the model answers given in your workbook.

RESEARCH ASSIGNMENT 3

Briefly compare the functions of the radiation protection regulatory authority in your country with the functions described in this module.  For this assignment, you will need a copy of your the radiation protection legislation and regulations of your country and/or state.  Your response should be no longer than 600 words and should be written in your workbook.

 4.RESOURCES FOR INTERVENTIONS

Where a radiation source is beyond the control or responsibility of the regulatory authority and licensees and registrants, the national infrastructure must provide the resources for interventions that will ensure adequate protection of the population.  Such interventions may be needed to mitigate hazards due to radiation accidents, natural sources, or residual sources from past practices.  In these cases, resources include the provision of trained personnel, monitoring and analytical equipment, communications and transport.

Intervention must be planned as part of the normal operations of the national infrastructure.  All types of situations that could lead to the need for intervention should be considered.  For each type of situation, plans must be made for the most appropriate response and agreed by all concerned.  To confirm their suitability for the situation, these plans must be exercised regularly and any necessary changes should be made to the plans to correct deficiencies found in the exercises.  Emergency plans must also be revised and updated to account for changes in personnel, policies and public expectations (see Figure 7).

Figure 7

Intervention Planning

Interventions for large-scale public hazards will most likely require the involvement of several government organizations.  It is important that the overall control is clearly defined and the responsibilities of each organization are clearly allocated within the national infrastructure, e.g. one organization may provide emergency radiation monitoring after a radiation accident, another organization may analyse and advise on intervention, and yet another organization may implement the intervention.

 5.EDUCATION, TRAINING AND INFORMATION

National infrastructures should support education and training as well as encouraging the exchange of information on the subject of radiation protection (see Figure 8).  Education and training should be supported to maintain an adequate standard of expertise in the country.  This education and training should extend to all levels of expertise, from simple workplace training to post-graduate study.  Such education and training may be provided by other countries or through organizations such as the International Atomic Energy Agency (IAEA).  Visits to radiation protection authorities of other countries can provide valuable experience, as well as education and training.

Figure 8

Education, Training and Information

Contacts with international organizations, national infrastructures of other countries, and between organizations within the infrastructure itself should be encouraged to ensure the flow of information on radiation protection.  A great deal of useful information on radiation protection is available from international organizations such as the ICRP and the IAEA and use of this information should also be encouraged.  Exchanges of information within the national infrastructure and with other countries will increase the efficiency of the infrastructure and help to ensure a uniform approach to radiation protection.

The flow of information from the national infrastructure to the public is also important.  The infrastructure should include means of informing the public on radiation protection issues directly or through the information media.  Public representatives (e.g. local politicians) should also be informed, to facilitate the governmental processes of setting priorities and allocating resources for radiation protection.

Contacts with the public should also involve the flow of information FROM the public TO the national infrastructure.

This is particularly important when policies are being established by the national infrastructure.  Whenever this occurs, the public's expectations for safety and protection should be taken into account.

 6.FACILITIES AND SERVICES

Some of the facilities and services necessary to maintain an adequate standard of radiation protection in the workplace are beyond the resources of most licensees and registrants.  The following facilities and services represent the minimum that should be provided by the national infrastructure:

• Facilities for intervention.

• Personal dosimetry services.

• Environmental monitoring services.

• Facilities for the calibration of radiation measuring devices.

• Radioactive waste storage and/or disposal facilities.

This is illustrated in Figure 9.

Figure 9

Facilities and Services

Provision of centralized facilities and services helps to ensure the effectiveness and uniformity of services across the range of authorized practices.  These centralized services should also provide central data bases such as national registers of occupational exposure records and information on radiation equipment.

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SELF-CHECK 5

Now see how much you have understood by answering the following questions in your workbook:

1. When would it be necessary for the national infrastructure to provide resources for intervention?

2. What are the two methods by which expertise in radiation protection may be maintained?

3. Why is the exchange of information within the national infrastructure and with other countries important?

4. What are the facilities and services which must be provided by a national infrastructure?

Now check your answers with the model answers given in your workbook.

KEY POINTS

0. The national infrastructure for radiation protection should provide the basic management and organization to ensure adequate radiation protection standards in a country.

1. The national infrastructure should consist of a legal framework; a regulatory authority; resources for interventions; education, training and information resources; and facilities and services.

2. The legal framework should provide regulations which make requirements for radiation protection and assigns responsibilities for meeting those requirements.

3. The regulatory authority should enforce the legal requirements.

0. The provision ofresources for interventions should ensure suitable response to situations beyond the capabilities or responsibilities of licensees and registrants or other organizations.

1. The provision ofeducation, training and information should maintain the national expertise and inform both organizations and the public on matters relating to radiation protection.

0. The provision offacilities and services beyond the resources of licensees and registrants should ensure that suitable safety standards are met and maintained.

1. Regulations require the possession, use and transfer of sources of radiation to be authorized by the regulatory authority.

2. Authorizations may be in the form of licences or registrations.

3. A licence is required where the safety of the practice is mainly dependent on the person conducting the practice.

4. Aregistration may be issued where the safety of the practice is ensured by the design of facilities or equipment, operating procedures are simple to follow, training requirements are minimal and few problems are known to be associated with the practice.

5. A practice, or a designated source within a practice, may be exempted from the requirements of the regulations if the radiation risk from the practice or source is of no concern to the regulatory authority, the collective radiological impact is sufficiently low and it is inherently safe.

6. Licensees, registrants and employers have responsibilities to establish objectives for radiation protection and to develop, implement and document a radiation protection program to ensure compliance with the regulations.

7. Radiation Protection Officers (RPOs) have the responsibility to provide technical advice, facilitate compliance with the regulations and arrange appropriate radiation protection training.

8. Workers have a responsibility to follow the rules and procedures for radiation protection, and to cooperate with the radiation protection program.

9. An independent regulatory authority must be established to ensure compliance with the regulations.

10. The functions of the regulatory authority are to assess applications for the authorization of practices or sources, inspect practices to confirm that regulations are being met, enforce compliance with regulations where necessary and encourage the development of a safety culture.

11. The general duties and responsibilities of inspectors are: to enter premises for the purpose of assessing compliance with regulations; to search premises for evidence of non-compliance; to examine records of a practice or source; to interview licensees, registrants, employers and workers regarding safety related issues; to test equipment, collect samples and analyse materials to assess compliance; to order the suspension of any practice, remove any source, or require any action to be undertaken to ensure adequate protection and safety where the inspector considers that there is an unacceptable hazard associated with a practice or source.

12. The following facilities and services should be provided by the national infrastructure to ensure that suitable safety standards are maintained; facilities for intervention; personal dosimetry services; environmental monitoring services; facilities for the calibration of radiation measuring devices.; radioactive waste storage and/or disposal facilities.

FINAL ASSIGNMENT

This assignment is to be completed under test conditions, so spend some time now studying this module.  Then contact your supervisor to arrange a suitable time to complete this assignment.

GLOSSARY OF TERMS

DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 3.2

Prepared by Safety and Radiation Science

Australian Nuclear Science and Technology Organisation

With the assistance of consultants in Australia and the United Kingdom, and IAEA experts from Korea, Mongolia, New Zealand, the Philippines and Thailand

Use of this material, acknowledging the IAEA and ANSTO as the source, is permitted for non-profit training

Acknowledgements

This material has been prepared by an extensive team of international experts from the fields of radiation protection, medicine and education.  It is not possible to mention by name all those who have had an input to; the initial planning; writing modules and workbooks; preparing and marking final assignments; proof-reading and editing; using the material and providing valuable feedback; organizing and hosting review meetings; and general administrative support.

The support of ANSTO management in providing the resource and encouragement is gratefully acknowledged.

The enthusiastic support of IAEA Technical Officers has been vital to the success of this project.

PART 3

RADIATION PROTECTION INFRASTRUCTURE

MODULE 3.2

TRANSPORT AND STORAGE OF RADIOACTIVE MATERIAL

CONTENTS

OVERVIEW

During the use of radioactive material, safe transport and storage of such material is often necessary.  In order to minimise radiation exposure during transport and storage, the principles of radiation protection are applied.  This module discusses the application of these principles in order to control the radiation hazards from the transport and storage of radioactive material.

The international transport of radioactive material is subject to controls set out by the International Atomic Energy Agency (IAEA) and these controls (referred to throughout this module as the ‘Transport Regulations’) are explained in detail in this module.  You will learn how safe packaging and transport can be achieved by the use of limits on radioactivity content and external dose rates.  You will also learn about the modes of safe transport and the responsible international organisations.

The desirable components of a radioactive material store are also described in this module, together with the design aspects, administrative controls, area and personal monitoring requirements for such a store.

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TIME ALLOCATION

It is estimated that you will need to spend eight to ten hours to complete this module before commencing the assignment.  The estimated time is for your guidance only.  Do not worry if you need more time to understand this module.

MATERIALS

You will need writing materials and your workbook while studying this module.

NOW YOU ARE READY TO START WORK

1. TRANSPORT REGULATIONS

At some point during the use of radioactive material, it may be necessary to transport the material either nationally or internationally.  Whether this transport is by road, rail, sea or air, there must be strict control to ensure the protection of persons, property and the environment.  For this reason, and to promote worldwide consistency of regulations for the transport of radioactive material, the International Atomic Energy Agency (IAEA) has developed a set of safety standards suitable for the transport of radioactive materials.  This module aims to introduce the most important aspects of these safety standards but does not cover all situations in detail.  Therefore, if you are intending to transport radioactive material, it is advised that you refer directly to the IAEA safety standards or to the regulations which apply in your country.

1.  Background

Radioactive material is classed as dangerous goods (Class 7) by the United Nations and is therefore subject to stringent methods of control.  To give some guidance on how this control should be administered, the IAEA has produced a set of safety standards known as the Regulations for the Safe Transport of Radioactive Material (ST-1).  These safety standards, commonly known as the Transport Regulations, are designed to comply with the Basic Safety Standards and as such, compliance with these Regulations is deemed to satisfy the principles of the BSS with respect to transport.

As is common with such complex topics as transport, the Transport Regulations are designed to be accompanied by two other IAEA documents which explain certain aspects in more detail.  Although the accompanying documents to ST-1 are still under review, the accompanying documents to its predecessor (IAEA Safety Series 6) are of some use when considering the current Transport Regulations (ST-1).  The first of these documents is the Explanatory Materialfor the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA Safety Series Number 7).  This document is designed to explain why these regulatory requirements are necessary.  It also explains the intent and rationale behind the Regulations and attempts to aid overall comprehension.

The second document Advisory Materialfor the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA Safety Series Number 37) is designed to provide information about the technical requirements of the Regulations and about methods and technology that may be employed to fulfil them.  In essence, this document provides information on how the Regulations may be met.

Together, these three documents provide a comprehensive guide to the basic transport safety standards expected by the IAEA (see Figure 1) and they should be considered together whenever transportation of radioactive material is necessary.

* currently being updated

Figure 1

The Basis of the Transport Safety Standards

It should be noted that, like the Basic Safety Standards, the Transport Regulations have no direct legal standing (i.e. they do not become national law unless incorporated into the legislation of that country).  However, in the case of international transportation, these regulations have been adopted by all international transport organizations and hence, they must be complied with for all international transport of radioactive material.

2. Objectives

The main aim of the Transport Regulations is to establish national and international standards of safety which provide an acceptable level of control of the hazards to persons, property and the environment that are associated with the transport of radioactive material.  This protection is achieved by combining four basic methods of control which can be summarised as follows:

0. containment of the radioactive material (This concerns the design and physical characteristics of a package, and the activity and nature of its radioactive contents);

1. control of the external radiation level (This concerns the maximum radiation level on the outside surface of a package and the package design to achieve this, the marking of a package with appropriate warning labels, and the safe stowage of a package during transport);

2. prevention of damage caused by heat (This concerns the maximum surface temperature of a package and instructions on stowage to provide for the safe dissipation of any heat generated);

3. for fissile material (e.g. reactor fuel), the prevention of criticality (i.e. prevention of a self-sustaining chain reaction).

An important underlying philosophy of the IAEA Transport Regulations is to ensure that the responsibility for ensuring safety during transport belongs mainly to the person or organization sending the material and, in this way, the contribution required from the carrier is minimised.  Transport industry workers are expected to treat radioactive consignments with care, but no more than accorded other dangerous goods.  The IAEA Regulations are therefore essentially directed to ensuring that levels of safety appropriate to the nature and amount of radioactive material being carried are built-in to the design of the package on the premise that the package could be involved in a transport accident.

3. Competent Authorities

The Transport Regulations require that a competent authority (or authorities) be appointed in each country to oversee the application of national and international transport regulations.  This responsibility is usually assigned to the national regulatory authority but in some cases, the regulatory authority may delegate this responsibility to another organization.

In essence, the competent authority is responsible for the following items:

• formulating national and international regulations;

0. issuing approvals for certain transport operations;

1. issuing approvals for certain types of package;

2. setting up a program to ensure compliance with the regulations; and

3. advising government, users and the industry of the regulations.

SELF-CHECK 1

Now see how much you have understood by answering the following questions in your workbook:

1. What is the main aim of the IAEA Regulations for the Safe Transport of Radioactive Material?

2. State the four basic methods of control by which the Transport Regulations provide radiation protection of persons, property and the environment.

3. What are the main functions of the competent authority?

Now check your answers with the model answers in your workbook.

2. GENERAL PROVISIONS OF THE TRANSPORT REGULATIONS

The Transport Regulations provide a number of general provisions which should be applied to all aspects of transport of radioactive material.  These provisions relate to the following topics:

0. Radiation Protection;

1. Emergency Response;

2. Quality and Compliance Assurance; and

3. Special Arrangement.

1. Radiation Protection

The Transport Regulations make the following provisions for the radiation protection of workers and members of the public:

0. A radiation protection program should be established for the transport of radioactive material to ensure compliance with the Regulations.  Inspection of the program should be performed by the competent authority to ensure that suitable standards are maintained.

1. The dose limits and ALARA principle from the Basic Safety Standards are incorporated in the IAEA Transport Regulations.  Hence, an effective system of radiation protection should be employed during transportation to ensure that all radiation doses resulting from the transport of radioactive material are as low as possible.

2. Workers should receive appropriate training concerning the radiation hazards involved and the precautions they should take in order to control their exposure and that of other people who might be affected by their actions.

3. The relevant competent authority should arrange for periodic assessments of the radiation doses to persons due to the transport of radioactive material to ensure that the system of protection and safety complies with the Basic Safety Standards.

4. Occupational exposures from transport activities do not require dose assessment or detailed monitoring programs if effective doses are unlikely to exceed 1 mSv per year.  Potential doses between 1 and 6 mSv per year require work place or individual monitoring, and if these potential doses are likely to exceed 6 mSv per year, individual monitoring should be conducted.

5. Distance should be used to control doses to transport workers and members of the public.  If the dose to transport workers is likely to exceed 5 mSv per year or the collective dose is likely to exceed 1 person mSv per year, the distances should be calculated to reduce this dose.

2. Emergency Response

The Transport Regulations require that emergency response arrangements are in place to ensure the protection of persons, property and the environment in the event of accidents or incidents during the transport of radioactive material.  These are usually provided by the national infrastructure and/or international organizations.

3. Quality and Compliance Assurance

The competent authority should ensure that the person or organization transporting radioactive material has in place a suitable quality assurance program.  This quality assurance program should be established and maintained in accordance with national or international standards for the design, manufacture, testing, documentation, use, maintenance and inspection of all packages and transport operations.  The competent authority is also responsible for assuring general compliance with the Transport Regulations.

4. Special Arrangement

In some cases, compliance with the Transport Regulations is impractical.  In these cases, special arrangements must be made with the competent authority before radioactive material can be transported.

SELF-CHECK 2

Now see how much you have understood by answering the following questions in your workbook:

1. What are the four general provisions of the Transport Regulations?

2. At what levels of potential dose should the following be conducted:

1. workplace monitoring?

2. individual monitoring?

3. Why are emergency response requirements part of the Transport Regulations?

4. Which organization is responsible for ensuring that suitable quality assurance programs are in place?

5. When are special arrangements necessary for the transportation of radioactive material?

Now check your answers with the model answers in your workbook.

3. SPECIFIC PROVISIONS OF THE TRANSPORT REGULATIONS

To minimize and control the hazard due to radioactive material during transportation, the Transport Regulations set out a number of specific provisions.  These provisions include the following:

0. the setting of exemption levels, below which material is considered not to present a radiological hazard and may be transported accordingly.

1. the setting of activity values to provide guidance on the amount and type of material which can be transported.

2. the setting of restrictions, known as activity limits, on the amount and type of material which can be transported in a particular package type.

3. details of package requirements and testing.

4. details of categorization labelling and documentation of packages.

1. Exemption Levels

Some consignments may contain levels of material which are not considered to present a radiological hazard and these levels are referred to as exemption levels.  In fact, the Transport Regulations define the term radioactive material for each radionuclide in terms of exemption levels for the specific activity (Bq g^-1) of radioactive material and the total activity (in Bq) of a consignment.  Hence, for consignments of material with specific and total activities below these exemption levels, the contents are not considered radioactive and are therefore exempt from the Transport Regulations.  Conversely, for consignments with specific and total activities above the exemption levels stated in the Transport Regulations, the contents are considered radioactive and should be transported as such.

The exemption levels for some common radionuclides are given in Table 1.

Table 1

Exemption Levels for Some Common Radionuclides

Radionuclide	Specific Activity Exemption Level (Bq g-1)	Total Activity Exemption Level (Bq)
Co-60	1 x 101	1 x 105
I-131	1 x 102	1 x 106
Tc-99m	1 x 102	1 x 107
Cs-137	1 x 101	1 x 104
Ra-226	1 x 101	1 x 104
U-238 (natural)	1 x 100	1 x 103
Am-241	1 x 100	1 x 104

2.  Activity Values

In order to minimize the hazard during transportation, the activity of radioactive material in a particular package design should be limited.  In the Transport Regulations, the maximum activity of a particular radionuclide allowed in different package designs is related to set activity values known as A₁ and A₂.  The value A₁ relates to the activity for special form radioactive material (i.e. an indispersible solid radioactive material or a sealed capsule containing radioactive material) and the value A₂ relates to the activity for radioactive material other than special form radioactive material.

The values are calculated on the basis of standard dosimetric models which consider a series of exposure routes for both internal and external hazards.  In each case, it is assumed that a person would be unlikely to remain at one metre from a damaged package for more than thirty minutes.

The values are given in the Transport Regulations for each radionuclide in units of terabecquerels (TBq) and Table 2 shows some A₁ and A₂ values for some common radionuclides.

Table 2

A₁ and A₂ Values for some Common Radionuclides

Radionuclide	A1 (TBq)	A2 (TBq)
Co-60	0.4	0.4
Tc-99m	8	8
I-131	3	0.7
Cs-137	2	0.6
Ra-226	0.2	0.03
U-238 (natural)	Unlimited	Unlimited
Am-241	10	0.001

SELF-CHECK 3

Now see how much you have understood by answering the following questions in your workbook:

1. What are exemption levels and how are they quoted in the Transport Regulations?

1. A geological rock sample containing natural uranium-238 is to be transported internationally.  If the total activity of the rock is 250 Bq and the total mass is 500 grams, would the package be exempt from the requirements of the Transport Regulations?

1. How do the Transport Regulations define the maximum permissible activity for each package design?

1. Which types of radioactive material do the activity values A₁ and A₂ relate to?

1. What is the activity value for

a)an unsealed source of iodine-131?

b)a sealed capsule of caesium-137?

Now check your answers with the model answers in your workbook.

3.  Package Types

In order to provide safe yet economical ways of transporting radioactive material, five types of packages are defined by the Transport Regulations.  These are as follows:

0. Excepted Packages;

1. Industrial Packages;

2. Type A Packages;

3. Type B Packages; and

4. Type C Packages.

By defining these package types, the Transport Regulations grades the level of hazard such that, as the hazard posed by the contents increases, the design requirements, standard tests and performance criteria become more stringent.

1.  Excepted packages

Excepted packages provide a means of transporting small quantities of radioactive material safely and at low cost.  It is assumed that in any accident situation involving an excepted package, all the radioactive contents could be released.  Hence, the contents of the package must be such that this release would present an insignificant radiological hazard.  To ensure that this is the case, the Transport Regulations define four specific types of excepted packages as follows:

0. Limited activity;

1. Small levels of radioactive material enclosed in or part of an instrument or manufactured article;

2. Packages containing articles manufactured from natural or depleted uranium or natural thorium; and

3. Empty packaging which previously contained radioactive material.

For the first two types, the Transport Regulations set out limits on the activity content of each package.  These limits depend on the physical state of the contents (solid, liquid or gas), whether the material isspecial form radioactive material and whether the material is part of an instrument or article (e.g. an americium-241 foil in a smoke detector).  In the case of manufactured instruments or articles, there are also limits on the activity for each particular item in a package as well as for a package as a whole.  These limits are shown in detail in Table 3.

Table 3

Activity Limits for Excepted Packages

For articles manufactured from natural or depleted uranium or natural thorium, an excepted package may contain any quantity of such material provided that the outer surface of the uranium or thorium is enclosed in an inactive sheath made of metal or some other substantial material.

In the case of empty packaging, the package may be transported as an excepted package provided that:

0. It is in a well maintained condition and securely closed.

1. The outer surface of any uranium or thorium in its structure is covered with an inactive sheath made of metal or some other substantial material.

2. The level of internal non-fixed contamination does not exceed:

3. 400 Bq cm^-2 for beta and gamma emitters and low toxicity alpha emitters; or

4. 40 Bq cm^-2 for all other alpha emitters.

5. Any labels which may have been displayed on it are no longer visible.

Whatever the type and activity of an excepted package, it is subject to stringent dose rate constraints.  For any excepted package, the dose rate at any point on the surface of the package must be no greater than 5 Sv h^-1.  In the case of components of manufactured instruments and articles, an additional constraint is applied such that the dose rate at 10 cm from any point of the unpackaged material must be no greater than 0.1 mSv h^-1 (see Figure 2).

Figure 2

Maximum Dose Rates for Excepted Packages

2.  Industrial packages

The permitted contents of industrial packages must belong to one of two categories of radioactive material – either they must be of low specific activity (LSA) or be surface contaminated objects (SCO).  LSA contents are further divided into categories I, II and III as follows:

0. LSA-I material is defined as being intrinsically safe such that a person cannot physically inhale enough material to exceed a committed effective dose of 50 mSv.  Natural uranium and thorium ores and concentrates, unirradiated natural or depleted uranium and thorium are included in this category.

1. LSA-II packages contain activity which is not usually uniformly distributed such as contaminated vials, hospital and decommissioning wastes.

2. LSA-III packages should contain only solid radioactive material and the activity must be uniformly distributed throughout.

The SCO category is also divided into two categories, I and II and these cover items which are not radioactive, but have radioactive material distributed on their surfaces.  Limits are placed on the fixed and non-fixed contamination on accessible surfaces and total contamination on inaccessible surfaces (see Table 4).

Table 4

Contamination Limits for Surface Contaminated Objects

 Note:  Surface contamination is averaged over 300 cm² or over the whole surface if the total area is less than 300 cm².

The different categories of industrial packages and examples of each are summarised in Table 5.

Table 5

Summary of the Categories of Industrial Packages

Industrial packages are also subject to activity limits depending on their physical state and how they are transported (see Table 6).

Table 6

Activity Limits for the Transportation of Industrial Packages

* Except in the case of air travel where the activity should not exceed 3000A₂

Industrial packages themselves are divided into three sub-divisions known as IP-I, IP-2 and IP-3.  These three packages differ in their ability to withstand minor mishaps (see Section 3.4.2 for test requirements) and provide a means whereby less hazardous material may be transported more cost effectively.  The Transport Regulations categorize the three industrial package types by their contents and whether the transport is under exclusive use (i.e. whether it is controlled directly by the organization sending the package).  This categorization is shown in Table 7.

Table 7

Industrial Package Types

Radioactive Contents	Exclusive Use	Not under Exclusive Use
LSA-1 Solid Liquid	IP-1 IP-1	IP-1 IP-2
LSA-II Solid Liquid	IP-2 IP-2	IP-2 IP-3
LSA-III	IP-2	IP-3
SCO-I	IP-1	IP-1
SCO-II	IP-2	IP-2

Dose rate limitations for industrial packages depend on how a package is transported (i.e. whether the package is transported by road, rail, sea or air) and whether it is transported under exclusive use.  Table 8 summarises the dose rate limits for industrial packages.

Table 8

Dose Rate Limits for Industrial Packages

3.  Type A packages

Type A packages are intended to provide a safe and economical means of transporting relatively small, but significant, quantities of radioactive material.  They are commonly used to transport radiopharmaceuticals and should contain activities no greater than A₁ for special form radioactive material, and no greater than A₂ for all other radioactive material.  They are designed to provide adequate shielding and containment under normal conditions of transport including mishandling such as falling from vehicles, being dropped during manual handling, being exposed to the weather, being struck by a sharp object, or having other packages and cargo stacked on top.  Under severe accident conditions, it is accepted that the containment may be breached with a partial loss of contents and the shielding may be destroyed.

A typical Type A package is shown in Figure 3.

Figure 3

A Typical Type A Package

The dose rate limits for Type A packages are similar to those for industrial packages.

4.  Type B packages

Type B packages are intended for the transport of radioactive material in excess of the Type A limits.  However, the design requirements for Type B packages are such that in the event of a severe accident, the shielding and containment are retained.  For this reason, Type B packages tend to be much sturdier than Type A packages and there are a variety of different designs available.  Different countries and organizations may have their own preferred design but a typical example of a Type B package is shown in Figure 4.

Figure 4

An Example of a Type B Package

Like Type A packages, Type B packages are subject to activity limits but rather than being fixed limits (except in the case of air transport), they depend on the ability of the package design to withstand accident situations.  Hence, each package design is required to undergo a number of tests with respect to the shielding, temperature, pressure and corrosion aspects to determine the limits on activity.

To ensure that the activity limits are documented and kept to, when Type B packages are to be transported, a safety submission must be presented to the competent authority for approval.  A certificate of approval will then be issued by the authority stating the limit on contents of the package.

If Type B packages are to be transported by air, upper limits are prescribed by the Transport Regulations.  These are as follows:

0. for low dispersible radioactive material  as authorized for the package design as specified in the certificate of approval;

1. for special form radioactive material  3000 A₁ or 100 000A₂ whichever is the lower; or

2. for all other radioactive material  3000A₂.

Again, the dose rate limits for Type B packages are similar to those for industrial packages.  Also note that Type B packages may either be unilaterally approved (a Type B(U) package), or multilaterally approved (a Type B(M) package).  Unilateral approval means that they are approved by the competent authority of the country of origin of the design only, while multilateral approval means that they are also approved by the competent authorities through, or into which, the consignment is to be transported.

5.  Type C packages

Like Type B packages, Type C packages are intended for transport of radioactive material in excess of the Type A limits.  However, Type C packages are subjected to more rigorous tests than Type B packages and have no activity limits for air transport.  In fact, Type C packages are usually used to carry levels of activity which fall into the upper limits of Type B packages and are designed such that they are unlikely to release radioactive contents or lose shielding integrity in the event of an accident including air transport accidents.

Like Type B packages, when transporting Type Cpackages,a safety submission must be presented to the competent authority.  The certificate of approval issued by the authority will then state the limit on contents of the package.

Note that Type C packages are subject to the same dose limits as Type A, Type B and industrial packages.

6.  Summary of package types

The different package designs are summarised in Table 9 with respect to activity and radiation level limits.

Table 9

Summary of Package Types

Note that special requirements exist for the transportation of fissile material and uranium hexafluoride but, due to their complex nature, they are not covered in this module.  For information regarding these requirements you should refer to the Transport Regulations directly.

SELF-CHECK 4

Now see how much you have understood by answering the following questions in your workbook:

1. What are the five types of package for radioactive material as defined in the Transport Regulations?

2. One hundred smoke detectors each containing a small sealed americium-241 source are to be transported in accordance with the Transport Regulations.  If the activity of each smoke detector is 37 kBq, what is the most suitable package type for this consignment and why? (Assume that, as Am-241 is primarily an alpha emitter, the dose rate criteria for all package types are met.)

3. A package containing combustible solid decommissioning waste is to be transported by road and under exclusive use to a national waste repository.  According to the Transport Regulations:

a)which is the correct category of industrial package for this consignment (LSA-I, LSA-II, LSA-III, SCO-I or SCO-II)?

b)what is the maximum activity for this package?

c)what type of industrial package should be used?

d)what is the maximum dose rate allowed on any external surface of the package?

4. What are the activity limits for Type A packages?

5. What is the main difference between the package designs for Type A and Type B packages?

6. Which type of package is subject to the most rigorous tests?

7. Which types of package require a safety submission to be made to the competent authority?

Now check your answers with the model answers in your workbook.

4.  Packaging Tests and Requirements

The Transport Regulations prescribe a number of tests and requirements relating to general package design and to package performance under both normal and accident conditions.  Note that special requirements apply to the transportation of fissile material and uranium hexafluoride but they are not detailed in this module.

1.  General requirements

According to the Transport Regulations, all packages (Excepted, Industrial. Types A, B and C) must meet a number of general requirements before they can be used to transport radioactive material.  These general requirements relate to the package’s physical properties (such as its mass, volume, shape and weight) as well as such things as the ease of decontamination of the package and its ability to withstand vibrations and variation in temperature.

2.  Tests to simulate normal transport conditions

To show that specific package designs are suitable for transporting radioactive material, the Transport Regulations prescribe a number of tests to simulate conditions which could occur during normal transport.  Each package type is then expected to pass the tests prescribed for its particular class.

As excepted packages are not expected to retain their contents after even minor mishaps, no specific tests are prescribed for this type of package.  This is also true of industrial packages in the IP-1 class.  However, all other package designs (IP-2, IP-3, Type A, Type B and Type C) are required to undergo tests relating to minor mishaps and must meet the test requirements for its class.  The four main prescribed tests are as follows:

0. A water spray test which consists of spraying the package with a water spray equivalent to a rainfall rate of 5 cm h^-1 for a duration of one hour to simulate bad weather conditions;

1. A free drop test (from 0.3m to 1.2m) which entails dropping the package onto a hard target from various heights to simulate being dropped during manual handling;

2. A stacking test which consists of applying a compression load equal to five times the weight of the package to simulate stacked packages; and

3. A penetration test from 1 mto simulate the penetration of the package by a sharp object which involves dropping a metal bar onto the weakest part of the package.

The integrity requirements for industrial packages in class IP-2 are such that the packages needs to be able to withstand the stacking and penetration tests only.  For industrial packages in class IP-3, however, the packages are expected to retain their contents during all minor mishaps and should therefore be able to withstand all four tests.

Type A packages are similar in that they are also required to retain their contents during normal transport conditions.  Hence, Type A packages are required to withstand all four tests plus a more stringent penetration test (from 1.7 m) which is prescribed in the case of packages with liquid or gaseous contents.

Type B and Type C packages are required to maintain their integrity even under severe accident conditions and are therefore subject to the above four tests plus other, more rigorous tests.

3.  Tests to simulate severe accident conditions

Excepted and industrial packages are not required to retain their contents under severe accident conditions and are therefore not required to pass any tests other than those listed above.  Type A packages have similar requirements except in the case of liquid or gaseous content, where an additional drop test is required from 9 m.

Type B packages are required to retain adequate shielding and containment in the event of a severe accident.  While they are not required to stand more than one accident, tests are performed in series with accumulative effects.  Mechanical tests for heavy high density packages involve an impact test followed by a penetration test.  These usually include dropping the package from 9 m onto a flat surface and then onto a mild steel bar from a height of 1 m.  For light low density packages, mechanical tests involve a drop test (from 9 m) followed by a crush test (dropping a 500g steel plate onto the package from 9m).

Type B packages are also subject to such tests as a thermal test (where the package is subjected to a temperature of not less than 800⁰C for not less than 30 minutes) and water immersion tests (where the package is immersed in water equivalent to a depth of 15 m for not less than eight hours).  Note that type B packages of activity greater than 10⁵A₂ are also required to undergo an enhanced immersion test (to a depth of 200 m for one hour).

Type C packages should undergo the same tests as Type B packages but with an enhanced thermal test (60 minutes test time rather than 30 minutes) and an enhanced immersion test (200 m for one hour) regardless of activity content.  Type C packages must also undergo an additional impact test at an impact speed of 90 m s^-1 and a puncture tearing test using a steel probe dropped from a height of 3 m.

4.  Pass requirements

All packages should comply with the general design requirements but the pass requirements for each test depend on the package type.  The integrity of the containment system and shielding should be visually determined on all packages undergoing tests and any faults or damage to the packages as a result of each test should be identified and recorded.  The Transport Regulations additional pass criteria for each test of package type as follows:

0. For IP-2 and IP-3 and Type A packages, tests should show no loss or dispersal of the radioactive contents and no loss of shielding which would result in a 20% increase in the radiation level at any external surface of the package.

1. For Type B package, the loss of radioactivity be no more than 10^-6 A₂ per hour.

2. For Type C packages, the loss of contents following these tests should not exceed the A₂ values (or 10 x A₂ for Krypton-85) in one week and the radiation level at 1m from the surface should not exceed 10 mSv h^-1 with maximum contents in the package.

5.  Summary of requirements

A summary of the requirements for each package type is shown in Table 10.

Table 10

Summary Of Requirements For Each Package Type

Key

 LG = For liquid or gaseous radioactive material only

 HHD = For heavy high density packages only

 LLD = For lightweight low density packages only

 HAP = For high activity packages (>10⁵A₂) only

SELF-CHECK 5

Now see how much you have understood by answering the following questions in your workbook:

1. The Transport Regulations stipulate a number of general requirements relating to the design of radioactive material packages.  Give three examples of what these general requirements relate to.

2. Name the four tests prescribed in the Transport Regulations which simulate minor mishaps during normal transport conditions.

3. Using Table 8, compare the test requirements for the following package types:

a)Excepted packages and IP-3 packages;

b)Type A packages and Type B packages; and

c)Type B packages and Type C packages.

Now check your answers with the model answers in your workbook.

5.  Categorization, Labelling and Documentation

Once radioactive material is placed in a suitable package type, the packages are then categorized and labelled on the basis of the radiation dose rates at the surface of the package and at 1 metre from the external surface.  This categorization and labelling is performed in order to control radiation doses to transport workers and members of the public.  In addition to these labels, appropriate documentation should accompany each consignment of radioactive material.

1.  Categories and labelling of packages

Packages are categorised as I-WHITE, II-YELLOW, III-YELLOW in accordance with the maximum surface dose rate (in mSv h^-1) and in accordance with the Transport Index (TI).  The Transport Index is a number assigned to a package to control the radiation risk and it is calculated as shown in Equation 1:

TI = R₁ x 100[1]

Where TI = the Transport Index

  R₁ = measured dose rate at 1 m from the surface in mSv h^-1

For example, if the dose rate from a package is measured as 0.01 mSv h^-1 at 1 m from the external surface, then the Transport Index is calculated as:

TI = 0.01 x 100 = 1

The categories of packages for labelling purposes are shown in Table 11.

Table 11

Categories of Packages for Labelling Purposes

Transport Index (TI)	Maximum radiation level on external surface (mSv h-1)	Package category
0	0.005	I-WHITE
>0 to 1	>0.005 to0.5	II-YELLOW
>1 to 10	>0.5 to2	III-YELLOW
>10	> 2 to  10	III-YELLOW (must be transported under exclusive use)

Apart from excepted packages, the relevant category of package should be indicated by placing the TI and category description onto labels fixed to the external surfaces.  This helps to easily identify the radiation levels from a package by the category on the label.  Where the Transport Index satisfies the condition for one category and the surface radiation level satisfies the condition for a different category, the package should be assigned to the higher (most restrictive) category.

Each label must also have the contents in terms of the type and activity of each radionuclide.  Excepted packages should bear the marking RADIOACTIVE on an internal surface which is visible on opening the package.

Figure 5 shows the labels for each category of package.

I-White

II-Yellow

III-Yellow

Figure 5

Labels for Each Package Category

2.  Documentation

The consignor (i.e. the person or organization sending the consignment of radioactive material) is responsible for including the following information in the transport documents accompanying the package:

• the proper shipping name and description;

0. the U.N. Class Number 7 (radioactive material are classed as dangerous goods (Class 7) by the United Nations);

1. the U.N. Number (a four digit number prefixed by UN);

2. the name or symbol of the radionuclides (for mixtures, the most restrictive or an appropriate description should be used;

3. a description of the physical and chemical form of the radioactive material;

4. the maximum activity of the radioactive material (in Bq) or mass of fissile material (in grams);

5. the total activity of the consignment as a multiple of A₂ (for industrial packages);

6. the Transport Index (for categories II-YELLOW and III-YELLOW);

7. the category of package (I-WHITE, II-YELLOW or III-YELLOW);

8. the Criticality Safety Index (for fissile material);

9. the identification mark (for each competent authority approval certificate);

10. a statement of contents; and

11. a statement of exclusive use (if material is shipped under exclusive use).

These documents should also include a declaration which states that the consignor has described the contents correctly and in all respects complies with the relevant international and national transport regulations.  This declaration should be signed and dated by the consignor and be included with the information given above.

Note that the person or organization which receives the package is referred to in the Transport Regulations as the consignee.

SELF-CHECK 6

Now see how much you have understood by answering the following questions in your workbook:

1. How is the category of a package for labelling purposes determined?

2. a)If the dose rate from a package is measured as 0.05 mSv h^-1 at 1 m from the external surface, what is the Transport Index?

b)If the dose rate at the surface of this package is 1.5 mSv h^-1, what is the category of package for labelling purposes?

c)Draw the label for this package category indicating the appropriate colours.

3. If the Transport Index of a particular package is 1 but the maximum dose rate at the surface of a package is measured as 0.55 mSv h^-1, what is the correct category for this package?  Explain your answer.

4. How should excepted packages be marked?

5. Who is responsible for ensuring that packages are correctly described in the relevant documentation and that they comply with the national and international regulations for the safe transport of radioactive material?

6. Give three examples of information which should be included in the documentation accompanying a consignment of radioactive material.

Now check your answers with the model answers in your workbook.

4.  STORAGE OF RADIOACTIVE MATERIAL

The storage of radioactive material is required for two reasons.  Firstly, to protect personnel and minimise the radiation hazard to people and the environment, and secondly for security reasons to prevent theft or unauthorised tampering.  This section describes the basic components of a radioactive material store that provide radiological protection for personnel but the security aspects of such a store are not discussed in detail.  Separate storage facilities should be provided for radioactive material and radioactive waste.  The stores used for these purposes should be kept locked at all times except when radioactive material is being transferred to or from it.

Note that the storage of nuclear materials such as reactor fuel and plutonium sources requires much more stringent conditions and is not considered here.

1.  Radiation Protection Principles

All material that is placed in a designated store should be prepared and contained in such a way that it can be handled and treated as sealed sources.  All sources in such a store can then be assessed in terms of external hazards only.  The exception to this is the production of radon and thoron gases and their daughter products where the emanation of these radionuclides may require monitoring for internal hazard assessment and gas removal using ventilation systems.

2.  Design of Facilities

During the design of a radioactive material store, the following aspects should be considered:

1. The segregation of the store from other hazards and stores.

2. The restriction of dose rates outside each store.  For example, dose rates could be restricted to

0. 10 Sv h^-1 at any point on its outside surface accessible to occupationally exposed persons.

1. 0.5 Sv h^-1 at any point on its outside surface accessible to members of the public.

Note that the location and occupancy time inside and outside of the store should be considered when determining dose rate limits.

3. The use of shielding for external walls and the use of walled areas inside to separate different types of sources.

4. The store may require a separate mechanical extract ventilation system discharging to the outside atmosphere.  If any of the contents of the store can give off significant amounts of a radioactive gas or vapour (such as radon or thoron), additional extract ventilation flow should be provided automatically when the store is opened.  Entry to the store should then be delayed for at least five minutes to allow removal of the radioactive gas or vapour.

5. The use of source handling and checking areas and, where leakage of source containers may occur, large trays to contain any spread of contamination.

3.  Administrative Controls

In order to minimise the hazards of storing radioactive material certain administrative techniques are required.  These include the following:

0. warning signs and labelling of source containers;

1. local rules ( including access controls and emergency procedures);

2. record keeping (including a source inventory and details of source movements); and

3. regular area and personal monitoring.

1.  Warning signs and labels

Each store should be provided with a clearly visible warning sign which:

0. displays the ionizing radiation warning (trefoil) symbol;

1. prohibits unauthorized entry; and

2. displays wording such as Caution – This store contains radioactive material.

Each container of radioactive material in the store should be provided with a label which displays:

0. the ionizing radiation warning  (trefoil) symbol;

1. the radionuclide(s) it contains;

2. the activity of the contents and date of measurement;

3. name and affiliation of the person responsible for the container; and

4. labels which comply with that described for transport (see Section 4.2) if the material is to be moved or transported out of the store.

2.  Local rules

Local rules for operation, entry and movement of material should be written and reviewed on a regular basis.  The local rules should give the name of the responsible person and should also outline the emergency procedures in the event of an accident or incident in the store.

3.  Record keeping

The licensee/owner/responsible person should maintain an up-to-date record of all radioactive material in the store.  This can be an inventory of current material and a record of their location at all times.

4.  Area and personal monitoring

The number and type of area radiation monitors employed will depend on the store contents.  For gamma emitting sealed sources a gamma detector with the display unit at the store entrance may be necessary and/or a hand held device to monitor the dose rates at each source location.  Sophisticated remote display or alarming monitors are usually not required for stores containing small quantities of radioactive material.  Contamination monitors, protective clothing and clean-up kits may be necessary for emergencies involving unsealed sources.

Personal monitoring may be required for persons entering the store on a regular basis but this depends on the dose rates and occupancy times.  Advice regarding the need for personal dosimetry should be sought from the radiation safety officer for the store or from the regulatory authority.

SELF-CHECK 7

Now see how much you have understood by answering the following questions in your workbook:

1. How should radioactive material be contained in a store?

2. Give two examples of factors which should be considered when designing a radioactive material store.

3. Give three examples of administrative controls for the operation of a radioactive material store.

Now check your answers with the model answers in your workbook.

KEY POINTS

0. The IAEA Transport Regulations have been developed to establish standards for the safe transport of radioactive material within and between countries.

1. Application of the Transport Regulations provides an acceptable level of control of the hazards to people, property and the environment which are associated with the transport of radioactive material.

2. Radiation protection may be achieved during transportation by the following methods:

3. containment of the radioactive material;

4. control of the external radiation level;

5. prevention of damage caused by heat; and

6. for fissile material (eg reactor fuel), the prevention of criticality.

7. The main function of a competent authority is to oversee the making of national (and international) transport regulations.  The competent authority is also responsible for the following items:

8. Issuing approvals for certain transport operations;

9. Issuing approvals for certain types of package;

10. Setting up a program to ensure compliance with the regulations; and

11. Advising government, users and the industry of the regulations.

12. The general provisions of the Transport Regulations relate to the following topics:

13. Radiation Protection.

14. Emergency Response.

15. Quality and Compliance Assurance.

16. Special Arrangement.

17. Exemption levels are levels of activity below which a material is not considered radioactive by the Transport Regulations and is therefore exempt from its provisions.

18. The Transport Regulations define the term radioactive material for each radionuclide in terms of exemption levels for thespecific activity (Bq g^-1) of radioactive material and the total activity (in Bq) of a consignment.

19. In the Transport Regulations, the maximum activity of a particular radionuclide allowed in different package designs is related to set activity values known as A₁ and A₂.

20. The value A₁ relates to the activity for special form radioactive material and the value A₂ relates to the activity for radioactive material other than special form radioactive material.

21. The five package types for the transport of radioactive material are: Excepted, Industrial, Type A, Type B and Type C.

22. Each package type has to comply with different requirements and tests as set out in the Transport Regulations.

23. The three types of package categories are I-WHITE, II-YELLOW, III-YELLOW.

24. These categories are based on the maximum surface dose rate and the Transport Index.  The Transport Index is a number assigned to a package to control the radiation risk and it is calculated by taking the measured dose rate at 1 m from the surface (in mSv h^-1) and multiplying by 100.

25. Each consignment must be accompanied by appropriate documentation which should include (where applicable) the following information: the proper shipping name, the U.N. Class Number; the U.N. Number; the name or symbol of the radionuclides; their physical and chemical form; the maximum activity of radioactive material; the total activity of the consignment in terms of A₂; the Transport Index; the category of package; the criticality safety index; any identification marks; a statement of contents and a statement of exclusive use.

26. The design of a radioactive material store should include the following features: segregation of the store from other hazards and stores, the appropriate dose rate criteria (taking into account the location and occupancy time); shielding, ventilation and the use of source handling and checking areas.

27. The administrative controls for a radioactive material store include warning signs and labelling of source containers, record keeping, local rules and regular area and personal monitoring.

FINAL ASSIGNMENT

This assignment is to be completed under test conditions, so spend some time now studying this module.  Then contact your supervisor to arrange a suitable time to complete this assignment.

 GLOSSARY OF TERMS

DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 3.3

Prepared by Safety and Radiation Science

Australian Nuclear Science and Technology Organisation

With the assistance of consultants in Australia and the United Kingdom, and IAEA experts from Korea, Mongolia, New Zealand, the Philippines and Thailand

Use of this material, acknowledging the IAEA and ANSTO as the source, is permitted for non-profit training

Acknowledgements

This material has been prepared by an extensive team of international experts from the fields of radiation protection, medicine and education.  It is not possible to mention by name all those who have had an input to; the initial planning; writing modules and workbooks; preparing and marking final assignments; proof-reading and editing; using the material and providing valuable feedback; organizing and hosting review meetings; and general administrative support.

The support of ANSTO management in providing the resource and encouragement is gratefully acknowledged.

The enthusiastic support of IAEA Technical Officers has been vital to the success of this project.

PART 3

RADIATION PROTECTION INFRASTRUCTURE

MODULE 3.3

SAFETY ASSESSMENT, REVIEW AND APPROVAL

CONTENTS

 OVERVIEW

This module provides you with information on safety assessment of a project in which radioactive substances or other sources of ionizing radiation are used for purposes such as research, medicine, industry, agriculture and education.  It is not intended to apply to major projects such as nuclear installations or major waste disposal facilities.

A safety assessment is usually needed at the design stage to assist in identifying the equipment and protection measures that are necessary.  In addition, a safety assessment of the project, as finally proposed, forms part of the safety submission that is needed to obtain safety approval for the project.  The precise requirements vary from country to country and also depend to some degree on the nature of the project and the scale of radiological hazard that it presents.  This module presents the general principles of safety assessment, preparation of a safety submission and the procedures and responsibilities for safety approval.

 

TIME ALLOCATION

It is estimated that you will need to spend eight hours to complete this module before commencing the assignment.  The estimated time is for your guidance only.  Do not worry if you need more time to understand this module.

MATERIALS

You will need writing materials and your workbook while studying this module.

NOW YOU ARE READY TO START WORK

5.  INTRODUCTION

Any practice involving the use of radioactive substances or other sources of ionizing radiation is subject to basic safety requirements to ensure the protection of people who could be exposed as a result of the practice.  The requirements applied in most countries derive from the Basic Safety Standards (BSS) for Protection against Ionizing Radiation and for the Safety of Radiation Sources, issued by the International Atomic Energy Agency, as Safety Series 115.  The standards specify requirements for notification of the Regulatory Authority of the intention to undertake a practice, and for authorization, either by registration or by licensing.

The legal body holding the authorization (i.e. the institution or company) is responsible for setting up and implementing the technical measures and the organizational systems that are needed to ensure that the necessary standards of protection are achieved.  Safety assessment plays an important part in the overall process of ensuring safety in any proposed practice.  In particular:

• an initial assessment assists in identifying the equipment and protective measures that are necessary, i.e. it has a key role in the design of equipment and procedures so as to ensure that dose limits will be met and that exposures will be as low as reasonable achievable; and

• in revised form, reflecting the equipment and procedures as proposed, the assessment is part of the safety submission that is required in order to obtain approval and/or authorization for the practice to commence.

6. DEFINITIONS

It is important that you understand the following terms and the meanings that are assigned to them for the purposes of this module:

• authorization – a permission granted by the Regulatory Authority in the form of a registration or a licence.

• institution - the premises, or area, in which a project is carried out.

• licence – an authorization granted by the Regulatory Authority on the basis of a safety submission and accompanied by specific requirements and conditions to be complied with by the licensee.

• licensee - the holder of a current licence for a practice or source who has recognized rights and duties for the practice or source, particularly in relation to protection and safety.

• normal radiation exposure - radiation exposure of persons expected to be received under normal operating conditions of a project.

• potential radiation exposure – radiation exposure that is not expected to occur but which could arise from events such as equipment failures or operator errors.

• practice - any human activity which increases radiation exposure.

• project - any work or operation in which radioactive substances or other sources of ionizing radiation are used.

• registrant - an applicant who is granted registration of a practice or source and has some recognized rights and duties for such a practice or source, particularly in relation to protection and safety.

• registration - a form of authorization for practices of low or moderate risks whereby the legal person responsible for the practice has, as appropriate, prepared and made a safety submission for the facilities and equipment to the Regulatory Authority.  The practice or use is authorized with conditions or limitations as appropriate.  The requirements for safety assessment and the conditions or limitations applied to the practice should be less severe than those for licensing.

• Regulatory Authority - an authority designated, or otherwise recognized, by a government for regulatory purposes in connection with protection and safety.

• Safety Advisor – a person or group with the necessary range of expertise to review a safety submission and grant approval for a project.

• safety approval – an internal consent for the project to proceed.

• safety arrangements – administrative arrangements and operational procedures to ensure safe operation.

• safety assessment - the analysis, appraisal and evaluation of the safety of a project.

• safety features - precautions, methods or devices proposed for the project which are designed to provide safety and to minimize the possibility of occurrence of accidents during its operation.

• safety review – an internal but independent review of a safety submission, including a safety assessment, by a suitably qualified and experienced person.

• safety submission - a written proposal describing the details and the safety features of a proposed new project, or significant changes to an existing project; the submission includes the results of the safety assessment.

• source - anything that may cause radiation exposure, such as by emitting ionizing radiation or releasing radioactive materials or substances.

7. WHEN IS A SAFETY ASSESSMENT NEEDED?

In general, any practice involving the use of radioactive substances or other radiation sources will be subject to requirements for notification of the regulatory authority and, in some cases, authorization.  In any such case a safety submission will need to be made to the regulatory authority and this will need to include a safety assessment.  However, where the quantity of radioactivity or the size of radiation source involved in a practice is very small, the practice might be exempted from the requirements for notification and authorization.  In this case, it is unlikely that a safety assessment would be needed.

IAEA Safety Series 115 specifies exemption levels for a large number of radionuclides.  These are expressed both as activity concentrations and as total activities held on the premises.  Some examples of the exemption levels are shown in Table 1.

Table 1

Examples of Exemption Levels

If either the concentration or the total quantity is less than that specified, then the practice may be exempted from the regulatory requirements.  If both values are exceeded, then regulation and safety assessment will be needed.

For other radiation sources, such as X-ray equipment or electronic devices, the source may be exempted from regulation provided that is of an approved type and that the dose equivalent rate at a distance of 0.1 m from any accessible surface of the equipment does not exceed 0.1 Sv h^-1.

SELF CHECK 1

Now see how much you have understood by answering the following questions in your workbook.

1. What is the role of safety assessment in a project involving sources of ionizing radiation?

2. When is a safety assessment required?

3. A project will make use of the radionuclide phosphorus-32 (P-32) in liquid solution at a concentration of about 10⁴ Bq/g.  The total quantity of liquor that will be handled at any time will be about 10 litres (i.e. 10 kg).  Under the IAEA Basic Safety Standards, will the project need to be notified to the Regulatory Authority and a safety assessment undertaken?

Now check your answers with the model answers in your workbook.

8. WHO IS RESPONSIBLE FOR WHAT?

In general, the Head of an Institution is responsible for all activities, projects and operations on a site, and for their safety.  Responsibility for particular projects may be delegated to a Head of Project and it is an important safety principle that the responsibility for safety rests with the person having authority over the project.  The Head of Project is therefore responsible for making the safety case for the project and this will include the safety submission and the supporting safety assessment.  The Head of Project would normally be assisted in this process by safety experts.

In order to provide an independent check on safety matters, the Head of the Institution will appoint a person or a group as Safety Advisor.  The role of the Safety Advisor is to undertake independent review of safety submissions and to advise the Head of the Institution on the granting of safety approval for a project.  This procedure ensures that the interests of safety are not subordinate to those of the project.  This internal safety review is undertaken and any concerns are addressed before the safety submission is made to the regulatory authority.

9. THE SCOPE AND CONTENT OF A SAFETY ASSESSMENT

As discussed earlier, safety assessments are required at different stages of a project.  An initial safety assessment is made in the early stages of a project to assist in identifying the safety provisions needed.  This will include:

• An analysis of the proposed equipment and operations;

• Establishment of design criteria in terms of radiation dose rates and allowable contamination levels in and around the facility;

• Characterisation of the sources of radiation that will be used and of the radiation dose rates and levels of radioactive contamination that will arise;

• Identification of potential accident or fault conditions that could lead to radiation exposure of workers in the institution or to releases of radioactivity to the environment, with the potential for exposure of members of the public;

• Estimation of the expected exposures from normal operations and, in broad terms, of the magnitudes of potential exposures from fault or accident conditions.  Here, account should be taken of external factors that could affect safety, such as loss of power or other services, and damage due to external occurrences such as seismic events.  These assessments should consider the impacts both on workers within the institution and on members of the public; and

• Confirmation that the safety equipment proposed (e.g. shielding, containment, ventilation and monitoring systems) is sufficient to meet the design criteria.

On large projects, this is an iterative process as the design of the project and the operational procedures develop.

The final safety assessment presents a review and analysis of the safety of the project, as finally proposed, to form part of the safety submission.  This will cover a similar scope to the initial safety assessment but the emphasis will be on demonstrating that the project can be operated safely.  For this purpose it needs to show that, with the safety equipment and safety arrangements provided:

• The radiation doses received by operators and others will be within the appropriate dose limits and be as low as reasonably achievable

• The risks from potential exposures are at acceptable levels.

SELF CHECK 2

Now see how much you have understood by answering the following questions in your workbook.

1. Who is responsible for the safety of a project?

2. What is the scope of a safety assessment?

3. Why is it necessary to undertake an independent safety review?

Now check your answers with the model answers in your workbook.

10. THE SAFETY SUBMISSION

An application for approval and/or authorization for a project must be supported by a safety submission.  This will normally be the responsibility of the Head of the Project, with appropriate specialist input.  The submission will include the results of the safety assessment.

You should therefore ensure that a safety submission is prepared whenever:

• a new project is proposed in your institution in which ionizing radiation and/or radioactive substances are used, or

• significant changes to an existing project are proposed.

As discussed earlier, the preparation of a written safety submission should be the responsibility of the person who will be in charge of the new project or of the revised existing project.

This ensures that this person:

• appreciates, and considers, all of its relevant safety aspects;

• takes account of, and implements, all the necessary features required to ensure safety; and

• implements the appropriate precautions to minimise all potential risks.

The person preparing the submission may not be skilled in evaluating all the potential hazards arising in the project and in such case will need to consult with and obtain assistance from other persons having appropriate expertise.

11. CONTENTS OF A SAFETY SUBMISSION

This module is concerned with projects in which ionizing radiation and/or radioactive substances are used so that the main consideration will inevitably relate to radiological hazards and provision of adequate radiation safety.  However you must not forget that other hazards may often be associated with such a project and these could include chemical hazards, electrical hazards, physical and mechanical hazards, microbiological hazards and non-ionizing radiation hazards.  Therefore it is essential that the safety submission also considers these other hazards in addition to the radiological hazards.  

A safety submission may be generic or specific. A generic safety submission will generally be suitable for a project of a type which has been conducted on a number of occasions previously, where the radiological hazards are well understood and the normally implemented safeguards have been shown to be effective.

A specific safety submission is required for a project where some, or all, of the radiological conditions are unique and there may be hazards that are not well understood.  A specific safety assessment may not need to consider in detail those aspects that have been considered previously in a generic safety assessment for a similar project.

The various items that need to be addressed in a written safety submission include:

• information stating when and where the project is to be carried out and its anticipated duration;

• a detailed description of the project supplemented by appropriate sketches, drawings, photographs and operations manuals;

• full details of the ionizing radiation sources and radioactive substances to be used in the project;

• a list of persons to be engaged in the project, their roles and responsibilities, and details of their relevant training and experience in the use of ionizing radiation and radioactive materials;

• details of operating and maintenance procedures;

• full details of the safety equipment, including items such as shielding, ventilation, containment and monitoring systems;

• a statement of previous safety experience with similar types of projects;

• statements of compliance with local radiation safety rules and with state/national legislative requirements for radiation control;

• the results of the safety assessment;

• details of the emergency equipment to be provided in anticipation of accidents and emergencies;

• the contingency plans for accidents and emergencies which may be anticipated in the project;

• the quality assurance program to be implemented in the project, including the arrangements for control of modifications to equipment or procedures (proposals for modification will need to show that safety will not be compromised either during or as a result of the modifications); and

• the procedures to be adopted in terminating the project and returning local conditions to normal.

SELF CHECK 3

Now see how much you have understood by answering the following questions in your workbook.

1. What is the purpose of a safety submission?

2. Who is responsible for preparing the safety submission?

3. What are the broad classes of information that need to be included in a safety submission?

4. What arrangements need to made to control changes to equipment or procedures?

Now check your answers with the model answers in your workbook.

12. THE REQUIREMENT FOR INDEPENDENT SAFETY REVIEW

As discussed above, internal independent safety review by the Safety Advisor is an important part of the process of assuring safety.  Its purpose is to verify the findings of the safety submission, and to confirm that the proposed project can be carried out safely and that it presents minimal risk to persons engaged in the project and to any other persons.

In summary the purpose of a safety review is to make a critical examination of:

• the safety logic inherent in the submission, including the criteria applied to the design of the project;

• the adequacy of the safety assessment, particularly in relation to the identification and quantification of radiation hazards;

• quality and extent of the radiation protection and safety provisions proposed for the project; and

• the acceptability of the estimated radiological impacts.

When satisfied that the proposed project can be carried out safely the Safety Advisor notifies the Head of the Institution.

13. SAFETY APPROVAL

It is essential for a safety approval to be granted before the project commences.  This is granted by the Head of the Institution on the advice of the Safety Advisor following review of the safety submission.  It is essential that the Safety Advisor is independent of the project, has appropriate knowledge and safety experience and has no direct involvement in the operation of the project.

In a specific institution the Safety Advisor may be:

• the Radiation Safety Officer (RSO) of the institution; or

• the Radiation Safety Committee (RSC) for the institution, if one exists.

If you are the RSO for your institution, and act as the Safety Advisor for a safety submission, you may sometimes find that a particular aspect of a safety submission is outside your knowledge and expertise.  In this case you should take advice on this aspect from a person(s) with the appropriate knowledge and expertise.  However you should ensure that such persons are completely independent of the proposed project.

One advantage of utilising the RSC of an institution to act as the Safety Advisor for a safety submission is that the committee has a number of members so that there may be wider overall knowledge and expertise available.

It is good practice for the RSO of the institution to examine the safety submission before it is formally assessed by the RSC.  This gives the RSO an opportunity to ensure that the safety submission is complete and that no relevant item has been overlooked.  If the RSO finds that a relevant item has been overlooked this should be brought to the attention of the person making the safety submission to ensure that it is included before being formally assessed by the RSC.

It is essential to ensure that a safety submission is made to the Safety Advisor well in advance of the proposed date of commencement of the project in order to allow adequate time for its safety assessment.  It also ensures that there is no delay in commencement of the project if the Safety Advisor decides that additional material needs to be included in the submission, or that additional safety measures are required to enable the project to be carried out safely.  It also means that the safety arrangements can benefits from operating experience.

Safety approval for a project may sometimes be granted for a specified period of time and require that a new safety submission be made after that time has elapsed.  In some cases an annual re-submission may be appropriate.  This ensures that any subtle changes in the project, which may have potential detrimental safety effects, are not overlooked.

The Safety Advisor may require a safety submission to be submitted in a particular format. If an agreed pro-forma, or a checklist, is used this assists in ensuring that all relevant information is provided.  Appendix 1 is an example of a pro-forma that could be used to present a safety submission.  You should note that the layout and space available in each section of a pro-forma for your own use might need to be different from those shown in the example.  After considering the safety submission the Safety Advisor should provide in writing:

• notification of the results of the review of the safety submission, with a recommendation for approval, subject to conditions where appropriate; or

• a request for a resubmission of the safety submission together with additional information or expansion of some details.

The Safety Advisor may decide to use a pro-forma for this similar to that shown in Appendix 2.

For some projects it may be necessary to submit a safety submission, not only to for independent review within the institution, but also to the Regulatory Authority of your State or Country.  A simple example of this is a project where your institution proposes to use a radioisotope tracer technique to trace the movement of sand on the bottom of a river or bay, outside its premises.  In such a case, the Safety Advisor would review the safety submission in the normal way and, if the Safety Advisor recommends its acceptance, the Head of the Institution would then submit it to the Regulatory Authority.  The Head of the Institution would not then grant safety approval until they have had approval from the Regulatory Authority.

SELF CHECK 4

Now see how much you have understood by answering the following questions in your workbook.

1. What is the purpose of independent safety review?

2. Who is responsible for undertaking a safety review?

3. How are the results communicated?

4. At what stage of a project is safety approval required?

5. Who grants safety approval for operation?

6. What is the approval procedure for projects that require regulatory approval?

Now check your answers with the model answers in your workbook.

KEY POINTS

• Authorization is the permission granted by the Regulatory Authority in the form of a registration or a licence

• Safety assessment means the analysis, appraisal and evaluation of the safety of a project.

• A safety submission is a written proposal describing the details and the safety features of a proposed new project, or significant changes to an existing project; the submission includes the results of the safety assessment.

• Safety review is an internal but independent review of a safety submission, including a safety assessment, by a suitably qualified and experienced person.

• Safety approval is an internal consent for a project to proceed.

• A safety assessment is needed for any practice that requires Regulatory authorization.

• The Head of an Institution is responsible for all activities, projects and operations on site, and for their safety.

• The Head of a particular project is responsible for making the safety case for the project.  This includes preparing the safety submission and the supporting safety assessment.

• The role of a Safety Advisor is to undertake the independent review of the safety submissions and to advise the Head of the Institution on the granting of safety approval for the project.

• The Safety Advisor may take the form of a Radiation Safety Officer (RSO) or a Radiation Safety Committee (RSC).

• Written safety submissions should include a safety assessment which covers both radiological and general safety.

• The various items that need to be addressed in a safety submission are covered in Section 7 of this module.

• Independent safety review is required to verify the findings of the safety submission and to confirm that the proposed project can be carried out safely and that it presents minimal risk to the persons engaged in the project and any other persons.

• The RSO of an institution should examine the safety submission before it is formally assessed by the RSC.

• The Head of an Institution is responsible for granting safety approval for a project based on the advice of the Safety Advisor following review of the safety submission.

• Where safety approval is required by the Regulatory Authority, the Safety Advisor will firstly review the safety submission as usual and if the Safety Advisor recommends it acceptance, the Head of the Institution will submit the submission to the Regulatory Authority.  The Head of the Institution would not then grant safety approval until they have had approval from the Regulatory Authority.

FINAL ASSIGNMENT

This final assignment is to be completed under test conditions, so spend some time now studying this module.  Then contact your supervisor to arrange a suitable time to complete the assignment.

 GLOSSARY OF TERMS

 APPENDIX 1

SAFETY SUBMISSION

Number..........

Date…........

Submitted to (name of safety assessor) for Safety Approval.

1. Name of Organisation/ Person Requesting Safety Approval

 …................................................................................………

2.Name of Project

..................................................................................……….

3.Location of Project

..................................................................................……….

4.Proposed Date of Commencement, and Duration, of Project

..................................................................................……….

5.Description of the Project (Where necessary attach details on additional sheets)

.....................................................................................................................................................................................................................................................................................................................................................................................................................................................................................………………………………………………………………

6.Description of Additional Items (eg. sketches, drawings, photographs, operating manuals) Included as Attachments

............................................................................................................................................................................................................................................................................................................................................

7. Full Details of Ionizing Radiation and Radioactive Substances to be Used in the Project

...............................................................................................................................................................................................................................................................................................................................................................................………………………………………………………………………….....................................................................................................................................................................................................

7. List of Persons to be Engaged in the Project and Details of their Experience and Training in Use of Ionizing Radiation and Radioactive Substances

..........................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................……………………………………………….

9.Name of Person in Charge of the Project (Included in 8)

...............................................................................................................

10.Details of Compliance with Local Radiation Safety Rules and with Legislative Requirements for Radiation Control

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11.Radiological Hazards Likely to Arise in the Project

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12.Details of Safeguards to be Provided

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13.Estimated Normal Radiation Exposures Likely to be Received in the Project

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14.Maximum Credible Accident Likely to Arise in the Project, Resultant Hazards and Potential Exposures

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15.Emergency Equipment to be Provided

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16.Details of Project Contingency Plans

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17.Operations Being Conducted in the Same Building or on Same Site which, if the Malfunction, may cause Possible Safety Problems for the Project

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18.Possible Effects of the Project on the General Public and the Environment During Normal Operation

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19.Details of Project Quality Assurance Program

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20.Details of Non-Radiological Hazards (eg. chemical hazards, microbiological hazards, electrical hazards, physical and mechanical hazards and non-ionizing radiation hazards

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21.Procedures to be Adopted in Terminating Project and Ensuring Pre-Project Normal Conditions

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Signature of Person Making Safety SubmissionDate

............................………………………………………..…../…../…..

Position of Person Making Safety Submission

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 APPENDIX 2

SAFETY APPROVAL

To:Name (Insert name of person who made the safety submission)

..........................................................................................................................

Organisation (Insert name of organisation where project is to be carried out)

..........................................................................................................................

Your Safety Submission Number ... ....... relating to:

• (insert name of Project)

• to be carried out at (Insert site of Project)

• commencing on (Insert commencement date) for the period (Insert dates for duration of Project)

Has been*/not been* given Safety Approval to proceed.

The reasons for withholding Safety Approval are as follows

............................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................................

You are therefore required to provide a revised safety submission which takes account of the above items.

(NOTE: *Safety Advisor to delete one of these as appropriate.)

Signature of Safety Advisor **:Date:

............................……………………………........…../…../…..

Title of Safety Advisor ***..................................……………………………

(NOTES:**Signature of the Safety Advisor if a single person eg. Radiation Safety Officer, or signature of the nominated person representing the Safety Advisor if it is a body such as a Radiation Safety Committee.

***Title such as Radiation Safety Officer, Radiation Safety Committee etc..)

DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 3.4

Prepared by Safety and Radiation Science

Australian Nuclear Science and Technology Organisation

With the assistance of consultants in Australia and the United Kingdom, and IAEA experts from Korea, Mongolia, New Zealand, the Philippines and Thailand

Use of this material, acknowledging the IAEA and ANSTO as the source, is permitted for non-profit training

Acknowledgements

This material has been prepared by an extensive team of international experts from the fields of radiation protection, medicine and education.  It is not possible to mention by name all those who have had an input to; the initial planning; writing modules and workbooks; preparing and marking final assignments; proof-reading and editing; using the material and providing valuable feedback; organizing and hosting review meetings; and general administrative support.

The support of ANSTO management in providing the resource and encouragement is gratefully acknowledged.

The enthusiastic support of IAEA Technical Officers has been vital to the success of this project.

PART 3

RADIATION PROTECTION INFRASTRUCTURE

MODULE 3.4

SAFE HANDLING OF RADIOACTIVE WASTE

CONTENTS

OVERVIEW

The use of radioactive material in industrial applications, research and medicine, as well as in the generation of nuclear power, gives rise to radioactive waste.  The importance of safe management of radioactive waste for the protection of human health and the environment has long been recognized and much experience has been gained in this field.  The International Atomic Energy Agency (IAEA) has developed and published standards and guidelines for radioactive waste management, which have been adopted internationally.  

Radioactive waste may occur in any of three physical forms (gaseous, liquid and solid) and range in concentration from trivial to highly concentrated.  Storage and disposal methods will vary according to the form of the waste, the radionuclides present, their concentration and radiotoxicity.

In this module, you will learn methods of managing radioactive waste safely in order to minimize hazards to people and the environment.  The module is designed for users of radioactive materials and addresses issues associated with waste classification, safe storage and appropriate processing to make the waste safe for disposal.   

Note that radiological hazards exist at all stages of waste handling and must be controlled.  You have learnt about methods of hazard control in previous modules and it will not be covered in detail in this module.  If you are unsure of appropriate techniques for the control of radiological hazards, you should revise the modules in Part 2 before commencing this module.

This module is not intended to address the requirements for waste repositories.

Before studying this module you should have completed the modules in parts 1 and 2.

 

It is estimated that you will need to spend 15 hours to complete this module before commencing the assignment.  The estimated time is for your guidance only.  If you would like to spend more time on the module to understand it more fully, then by all means do so.

MATERIALS

You will need writing materials and your workbook while studying this module.

NOW YOU ARE READY TO START WORK

14. Definition of Radioactive Waste

What do we really mean when we use the term “radioactive waste”?  The IAEA defines radioactive waste as  

“any material that contains or is contaminated with radionuclides at concentrations or radioactivity levels greater than ‘exempt quantities’ established by the competent regulatory authorities and for which no further use is foreseen.”

It should be recognized that this definition is purely for regulatory purposes, and that material with activity concentrations equal to or less than clearance levels is radioactive, although the associated radiological hazards are negligible.

Although a lot of waste can easily be identified as that “for which no further use is seen”, such as soiled tissues, care needs to be taken applying this definition when some of the waste may be recycled.  Some examples of recycling could be smelting contaminated scrap steel, radioactive sources that are too small for one application but could be used for another, or further use of uranium and plutonium from spent nuclear fuel.  It should be remembered that whether the radioactive material is recycled or disposed of as radioactive waste, it must be controlled as required by the national Regulatory Authority.

15. Radioactive Waste Management

    1. Objective

Radioactive materials are potentially hazardous if they are not controlled.  In addition to the controls on the use of radioactive materials, it is very important that the disposal of waste containing radioactive materials is also controlled.  Effective control of radioactive waste prevents unnecessary radiation exposure that may be harmful to people or the environment.  Since many radioactive materials have long half lives, some controls are required many years and even for many generations after the disposal.  Therefore, it is vital that in countries where radioactive materials are used, there is an ongoing national Radioactive Waste Management Programme to control the disposal of radioactive materials.

The internationally agreed objective of radioactive waste management is to deal with radioactive waste in a manner that protects human health and the environment now and in the future, without imposing undue burdens on future generations.

Radioactive waste in the environment can have an effect on organisms other than man.  However, since humans are among the most radiation sensitive organisms, safety standards that are sufficiently restrictive to protect humans will also protect other species.

The possible impact on the human health and environment of neighbouring countries should be taken into account by the waste management practices of a particular country.  The radiological impact on human health and the environment in other countries should be no more than those judged acceptable within the country of origin of the waste.  In fulfilling this duty, a country should take into account the recommendations of international bodies such as the ICRP and IAEA.  In particular IAEA Safety Series 67 “Assigning A Value To Transboundary Radiation Exposure” covers the potential effect of waste disposal processes on other countries.

2.  Waste Minimization

Good waste management minimizes the amount of radioactive waste generated, thus minimizing the hazard and maximizing the use of space in disposal facilities.

Waste minimization can be achieved in a number of ways.  Effective waste minimization typically employs a combination of more than one technique.  

• Delay and decay – Store short half life material to allow decay to levels below exemption levels.

• Dilute and disperse – provide sufficient dilution of liquid and gaseous wastes to below exemption levels.

• Review work practices to minimize quantities of radioactive material used where possible.

Do not allow unnecessary materials to go into areas where they are likely to become contaminated!  

3. A Radioactive Waste Management Programme

The control of radioactive waste is achieved by an effective waste management programme.  The programme, shown schematically in Figure 1, prescribes processes which may be required to ensure that waste can be disposed of safely.

Figure 1

Structure of a radioactive waste management programme.

You may find it useful to refer back to Figure 1 while studying the module.

Processing is divided into pretreatment, treatment and conditioning, all of which will be explained in section Basic Steps in Radioactive Waste Management.  Storage and transport may occur at any stage.

SELF CHECK 1

Now see how much you have understood by answering the following

1. What is the regulatory definition of radioactive waste?

2. What is the main objective of radioactive waste management?

3. What are 3 ways of minimizing radioactive waste?

4. Using the outline in your workbook, fill in the missing steps in radioactive waste management programme structure using the following word list.

Now check your answers with the model answers in your workbook.

16. Classification of Radioactive Waste

A suitable classification system for radioactive waste will help to plan the safe disposal of the material.  The classification includes the relatively simple classification of solid, liquids and gases and also whether the material is to be discharged to the environment or whether it is to be disposed of in a landfill or a specially built facility.  Liquid aqueous wastes are classified further according to concentration of radioactive material.  A more complicated classification by activity level and half-life gives the opportunity for specific guidance to be given on disposal of materials.

1. Environmental Discharges

Radioactive materials that are discharged directly into the environment and dispersed by it, are usually liquids or gases.  Solid waste is not discharged directly to the environment as it can be controlled at the point of production and placed in containers for decay or disposal to a suitable solid waste disposal site.  

Generally, environmental discharges are the effluent from a practice where radioactive materials are being used.  Examples of liquid effluent are radioactive materials disposed of down laboratory drains and the contents of discharge lines from industries such as phosphoric acid or rare earth extraction plants.  Examples of atmospheric releases are via the stacks of fume cupboards in which radioactive materials are being used or a licensed incinerator where combustible materials contaminated with radioactive materials are being burned.  Note that ash from incinerator stacks (solid material) could enter the environment unless it has been filtered out.

2. Specific Guidance For Liquid Waste

If highly contaminated liquids or liquids containing significant alpha activity are produced as waste, they are not suitable for discharge into the environment and are treated prior to disposal.  Section Basic Steps in Radioactive Waste Management will describe treatment methods.  Table 1 gives IAEA guidance on liquid waste.  Note that the activity concentrations in the table do not apply to liquids containing alpha emitting nuclides.

Table 1

IAEA classifications for radioactive liquid waste.

From “Handling and Treatment of Radioactive Aqueous Waste” IAEA TECDOC-654(1992)

* Releases require authorization (see Section AUTHORIZATIONS).

3. Solid Waste

Solid waste includes general lab waste, such as tissues and gloves, used filter units, source capsules from nuclear gauges and bulk solid materials, such as the residue from rare earth extraction processes or uranium mining.

Later in the module, you will learn how solid waste is managed and prepared for disposal.

4. International Waste Classes

Radioactive waste is classified in order to assist in the determination of appropriate disposal methods.

The IAEA have defined waste classes by their activity level and the amount of heat generated by the radioactive material.  The latest version is contained in the document IAEA Safety Series No 111-G-1.1 (1994).  The following figure shows a matrix of radioactivity levels and decay periods. There are three main areas in the matrix, Exempt Waste (EW), Low Intermediate Level Waste (LILW) and High Level Waste (HLW).  The LILW area on the matrix is subdivided into two areas according to radiological half lives (short lived (SL) and long lived (LL)).  

Figure 2

Waste classification system.

1. Exempt waste (EW)  

Exempt waste is waste excluded from regulatory control because its radiological hazards are negligible.  Regulatory Authorities establish exemption levels or clearance levels in terms of activity concentration and total activity.  IAEA BSS 115 Schedule I gives examples of clearance levels.

ACTIVITY

Find out how waste is classified as exempt in your country.  Discuss your findings with your supervisor.

2. Low intermediate level waste (LILW)  

Low intermediate level waste is typically divided into short-lived (SL) and long lived (LL) waste.

• Short lived waste (LILW-SL) does not contain significant levels of radionuclides with half lives greater than 30 years.  As a guide, activity concentrations of long lived alpha emitters may be limited to an average of 400 Bq g^-1 with a maximum package limit of 4000 Bq g^-1.  Disposal may be to a near surfacerepository.

• Long lived waste (LILW-LL)  contains long lived radionuclides in quantities that need a high degree of isolation from the biosphere.  Recommended disposal is in deep geological formations.  

The following are examples of materials which may be classified as LILW if the levels of total activity or radioactive concentrations are above the exemption levels.

• ion exchange resins and filter materials used to clean water at a nuclear power plant.

• research equipment from laboratories where radioactive materials are used.

• shoe covers, lab coats, cleaning cloths, paper towels, etc., used in an area where radioactive material is present.

• containers, cloth, paper, fluids, and equipment which came in contact with radioactive materials used in hospitals to diagnose or treat disease.

• filters from sampling devices used to test for airborne radioactive contamination.

• scintillation fluids in which filters from some sampling devices must be dissolved in order to determine the amount of radioactive material present.

• carcasses of animals treated with radioactive materials used in medical or pharmaceutical research.

3. High-level waste (HLW)

High level waste is waste that is sufficiently radioactive to require both shielding and cooling.  High level waste has concentrations of long lived radionuclides which exceed the limits for LILW-SL and the radioactive decay process produces heat at a rate above about 2kW m^-3.

HLW is generated during the reprocessing of spent nuclear fuel elements.  The waste contains significant levels of alpha emitters (mainly transuranics with long half-lives) and mixed fission products including radionuclides such as Caesium-134 and 137, Strontium-90, and Technetium-99.  The waste contains high levels of activity and generates enough heat of decay to require special cooling arrangements.

The waste classes and recommended disposal options are further described in Table 6 in the appendix.

SELF CHECK 2

Now see how much you have understood by answering the following

1. What characteristics of radioactive waste are used to classify it?

2. Why is liquid/gaseous waste treated differently to solid waste?

3. What is the level (in Bq g^-1) at which radioactive waste for each of the following radioisotopes is considered exempt waste in your country?

³H, ³²P, ⁶⁰Co, ⁹⁰Sr/⁹⁰Y, ^99mTc, ¹³¹I, ²⁴¹Am

4. How is low intermediate level waste subdivided?

5. What are the two main characteristics of high level waste?

6. Determine an appropriate waste classification for the following:

1. 100 TBq Co-60 source no longer required for a teletherapy unit.

2. 200 kBq I-131 in the discharge to the sewer of a nuclear medicine department.

3. The liquor resulting from reprocessing of nuclear fuel.

4. General lab waste containing material contaminated with tritium, Cs-134, Kr-85 and Co-60.

5. Contaminated lab coat after technician works with Tc-99m.  

Now check your answers with the model answers in your workbook.

17. International and national programmes

    1. The Role Of The IAEA

The IAEA has been assisting member states in the management of radioactive waste for around three decades.  An integrated waste management programme exists to help member states in the safe and effective management of radioactive waste.  The IAEA programmes are summarized in Table 2.  

The international exchange of technical information facilitated by the Agency’s waste management programme has led to the development of standards and criteria for the handling, treatment, conditioning, storage and disposal of radioactive waste of all classes.  As a result of these activities many Safety Series and Technical Reports Series documents have been published.

Table 2

Programmes established in the IAEA for Radiation Waste Management

2. International Co-operation  

International co-operation has been encouraged through the development of the Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management.

This convention is a legal instrument (effectively a law for the member states) requiring member states who sign the convention to undertake certain actions in regard to spent fuel and radioactive waste management.  The objectives of the convention is to achieve and maintain a high level of safety worldwide in the management of spent fuel and radioactive waste through the enhancement of national measures and international co-operation.  It is also aimed at preventing accidents with radiological consequences and mitigating their consequences should they occur during the management of spent fuel and radioactive waste.

3. National Framework For The Management Of Radioactive Waste

A national framework for waste management is necessary to protect the country’s citizens (and those in surrounding countries) from unnecessary exposure to radioactive materials.

The basic requirements of a national framework should be defined in the Regulatory Authorities regulations dealing with radioactive materials.  The following issues should be considered:

0. Setting levels of radioactivity or radioactive concentrations at which regulatory control is required.  (These levels may be taken from IAEA Safety Series documents.)

• Specifying public dose constraints which should be met in the disposal of radioactive waste.  (IAEA Safety series documents, such as IAEA Safety Standards Series No WS-G-2.3, “Regulatory Control Of Radioactive Discharges into the Environment”, give guidance on setting appropriate dose constraints.)

0. Having in place a system of authorizations for discharge, disposal and storage of radioactive waste.  (This should also include advice on waste treatment and disposal options.)

1. Specifying the monitoring requirements for the measurement of waste, and where necessary monitoring the environment.

2. The keeping of appropriate records of measurements and waste disposal.

3. Requiring a quality system to ensure the action taken and records kept are appropriate.

SELF CHECK 3

Now see how much you have understood by answering the following

1. What is the role of the IAEA in the management of radioactive waste?

2. What issues should you consider when establishing a national framework to deal with radioactive waste?

Now check your answers with the model answers in your workbook.

18.  AUTHORIZATIONS

Radioactive materials which are above the level subject to regulatory control are used under an authorization issued by the Regulatory Authority.  The storage, treatment and disposal of radioactive waste must also be subject to authorization.  The uncontrolled disposal of radioactive waste could expose large sections of the public, as well as occupationally exposed people.  Authorizations are important in order to ensure radioactive waste is dealt with safely and meets the goals of protecting human health and the environment.

IAEA recommendations on waste disposal and discharges into the environment are a good basis for the specific national requirements for authorization of radioactive waste discharges and disposals.  Typical requirements for authorizations and their compliance are discussed below.

1. Authorizations For The Storage Of Waste

Authorization to store radioactive waste is dependent on consideration of the physical security of the waste, the doses workers might be exposed to from the store, and the potential for accidental releases of the waste into the environment.  The authorization application should state the following as  minimum information.

0. The radionuclides to be stored.

1. The maximum activities to be stored.

2. The form of the waste, e.g. liquid scintillant, paper etc.

3. The storage period.

4. Details of the store and the containers the waste will be kept in.

5. The reason for storage.  For example, to allow decay or awaiting disposal.

The authorization will specify the radionuclides, their maximum activities and the period of time for which the waste can be stored on site.  The storage conditions may also be specified.

2. Authorizations For Waste Disposal

An authorization for the disposal of waste will specify the quantities of isotopes which can be disposed of during a certain time interval and will state the disposal site (e.g. landfill, incinerator, or waste processing plant).  The activity limit will probably be a simple fraction of the activity the disposal facility is allowed to receive for that period.

3. Discharges Into The Environment

Discharge of waste into the environment is a special case of disposal and requires very careful regulation.  The waste is moving away from the control of any authority or organization and into the biosphere with potential for intake by people directly or through the food chain.

The radiation exposure any person might receive from contact with the discharged material must not exceed dose limits, should be within dose constraints and be considered to be ALARA.  Although the radiation exposure to an exposed person is the limiting factor, for convenience the authorization to discharge to the environment will usually specify the maximum activity of the radioactive waste that can be discharged.  This may be in the form of the total activity discharged either per year, week, month or quarter.  To enable the Regulatory Authority to determine if the requested discharge is acceptable, the following information would be the minimum provided.

0. The radionuclides to be discharged.

1. Details of the type of waste and the method of discharge.  For example, “liquids from radioisotope laboratory drains discharged to a river via the site effluent line”.

2. The activity of the waste to be discharged and at what frequency.

3. Any dilution before or after it leaves the operators control.

4. An estimate of the potential public and any worker doses from the discharge.

5. An explanation why the waste should be discharged to the environment and not treated and sent for long term storage.  This could include a simple ALARA assessment.

Discharges of radioactive materials directly into the environment are usually liquids or gases.  The approved discharge levels specified in the authorization will be based on calculations of the radiation dose a person might receive from ingesting water contaminated by the waste, or from breathing the gas.  

1. Critical groups

When discharging liquid or gaseous waste into the environment, it is important to create a hypothetical group of people who will be the most likely affected by contact with radioactive materials.  The group of people is called the critical group.  By taking into consideration the radioisotope, its pathway into the body and the probability of the group coming into contact with the effluent, it is possible to predict a worst case scenario and calculate potential doses from the discharge.  The critical group is the group that would receive the greatest dose.  The product of the age related dose conversion factor (from IAEA BSS 115) and the intake for that age group will indicate what age group should be used as the critical group.  The effect of radiation on children is greater than that on adults, so the critical group is often children or infants.  The discharge limits are often based on the potential effect on children rather than adults.

Consider the following examples:

• Table 3 is used to determine the critical group when considering the disposal of tritiated water into a river from which drinking water is drawn.  Notice that the significant pathway is ingestion, therefore the data in the table refers to the ingestion of water.

Table 3

Determination Of Critical Group Due To Ingestion Of Tritiated Water.

* IAEA BSS No.115   

# Safety Report Series No.19

Study the table.  Which is the critical group?  You should notice that infants are the critical group since the product of the does conversion factor and the annual water intake is greatest for this group.

• Table 4 is used to determine the critical group when considering the disposal of gaseous I-131 through a stack and the inhalation pathway.

Table 4

Determination Of Critical Group Due To Inhalation Of Gaseous I-131.

* IAEA BSS No.115   

# ICRP 71

Study the table.  Which is the critical group?  You should have noticed that infants are the critical group since the product of the dose conversion factor and the inhalation rate is greatest for this group.

2. Determining discharge limits

In order to determine discharge limits, it is necessary to go through the following steps:

• Choose a model.

Usually a ‘worst case’ scenario, using relatively simple concepts, is chosen for screening purposes.  This will determine whether the discharge limits are realistic.  It may then be necessary to repeat the process using a more complex model requiring more data.

• Consider the most significant exposure pathways.

Exposure pathways include inhalation, external exposure and ingestion by direct drinking of water or indirectly through food.  Sometimes the significance of each pathway will be obvious.  It may depend heavily on whether the discharge is a single event or is continuous and expected to take place over many years.  Sometimes it is necessary to go through the calculation process to determine the significance.

• Determine the critical group.

You need to have decided the most significant pathway.  Then the critical group is determined as explained in Section 5.3.1.

• Collect appropriate data.

The data will include information such as total liquid discharge and water flow information for liquid discharges to a river, and information about the stack flow for gaseous discharges.  You will also need dose conversion factors and intake parameters.

• Apply a dose constraint.

The dose constraint will be provided by the Regulatory Authority.

• Use the above information to find out the maximum activity which can be discharged that would give a dose to members of the critical group equal to the dose constraint.

The following examples for liquid waste and gaseous waste show you how to determine discharge limits in a simple way.  Refer to Appendix 2 for the calculations.  Remember that here we have simplified the process in order to help you understand it.  Your objective is to understand how discharge limits are applied and what simplifications and assumptions have to be made.  The information in the Appendix will help this, but it is recommended that you work through the examples later, preferably after you have worked through the rest of the module.

LIQUID WASTE EXAMPLE

A laboratory wants to discharge tritiated water into a local river.  The river passes a water treatment plant where the water is taken, treated (with no further dilution) and supplied to the local town as drinking water.   

The laboratory must provide information about the proposed discharge to the Regulatory Authority who will then check the details and the projected doses to the public and any potentially exposed worker, prior to issuing an authorization and setting a limit on the discharge.

The tritiated water is disposed of through a sink drain into the site effluent, which has a discharge rate to the river of 10 000 m³ per week.

The river has a flow rate of 100 000 m³ per week.

The Regulatory Authority will determine an average weekly discharge limit which will comply with the annual dose constraint of 0.1 mSv.

To determine the discharge limit, we need to go through the six steps above.  The following table shows how to do this.

Table 5

GASEOUS WASTE EXAMPLE

A radioisotope production plant is seeking authority to discharge gaseous iodine-131 from its stack.  The plant must provide information about the proposed discharge to the Regulatory Authority who will then check the details, prior to issuing an authorization and setting a limit on the discharge.  The stack is 60m tall and the cross-sectional area of the stack at the point of release is 0.7 m².  The velocity of the discharge is 15 ms^-1.

To determine the discharge limit, we need to go through the six steps above.  The following table shows how to do this.

Note that both these examples are simple scenarios.  In reality, you will encounter discharges that have more than one radionuclide present and the limit that is imposed by the Regulatory Authority will include ALL radionuclides and ALL pathways.

3. Dose estimates

Potential doses to people as a result of planned or actual liquid or airborne discharges require estimating for the following reasons:

• To provide data to the Regulatory Authority when requesting discharge authorization.

• To ensure that doses are ALARA.

• To ensure that doses are in compliance with dose constraints.

To following list is the steps that need to be taken to estimate doses.

• Determine the nuclides and quantities discharged.

• Determine the pathways by which these nuclides may cause exposure.

• Determine the location and habits of the people who may be exposed.  If necessary, assume a ‘worst case’ scenario.

• Calculate  the doses from the individual nuclides in the discharge from each possible pathway.

• Sum the doses.

Appendix 3 has some simple examples of estimating doses.

Remember that, in reality, there are many more contributing factors to doses from both liquid and gaseous waste.  Radioactivity is deposited on the ground  and absorbed by plants and animals, usually from more than one radioisotope.  People are exposed to radioactive material that has been absorbed by grains and vegetables, from meat and fish and milk may be a contributing factor.    

If you need to take these factors into consideration, please refer to IAEA Safety Reports Series No. 19 “Generic Models for Use in Assessing the Impact of Discharges of Radioactive Substance to the Environment” (2001).

SELF CHECK 4

Now see how much you have understood by answering the following

1. What information should be provided to the Regulatory Authority when seeking authorization to:

1. discharge liquid waste?

2. store waste while waiting for disposal?

2. What does the term ‘critical group’ mean and how is it used?

3. List the 6 steps in determining discharge limits.

4. What is a screening calculation?

5. List some of the data required in order to calculate discharge limits for airborne effluent.

6. Why is it necessary to estimate doses from discharges to the environment?

7. List some of the limitations of simple dose estimation calculations.

Now check your answers with the model answers in your workbook.

19.  Basic Steps in Radioactive Waste Management

Effective management of radioactive waste considers the steps (shown schematically in Figure 3) in the radioactive waste management process as parts of a total system, from control of its generation through to disposal.  It should be noted that characterization, storage and transportation may take place between any of the waste management steps.

Figure 3 shows the three basic steps in processing radioactive waste prior to disposal.  The waste may be reclassified as exempt waste or removed from the waste stream for reuse at any stage.

Figure 3

Basic Steps of Radioactive Waste Management.

1. Disposal And Predisposal

Disposal is the final step in radioactive waste management in which radioactive waste is discharged into the environment or is placed into a disposal facility, without the intention of retrieval and without reliance on long term surveillance and maintenance.   

Discharge of liquid and gaseous effluents into the environment with subsequent dispersion (dilute and disperse) must be within authorized limits.  This is an irreversible action and is considered suitable only for limited amounts of specified waste.

Safety in a disposal facility is achieved by containment and isolation of suitably conditioned waste by a multilayer barrier system, either natural and/or engineered, to restrict the release of the radionuclides into the environment.  The barrier system is designed according to the disposal option chosen and the radioactive waste forms involved.

Predisposal is all the steps which have to be carried out prior to disposal of the waste.  The rest of section 6 will describe predisposal activities.

Radiological hazards exist in all predisposal processes.  Radiation protection practices must be in place to control personnel exposure to radiation and contamination.  Time, distance and shielding must be used to minimize worker’s exposure to radiation dose rates.  Contamination control methods, using containment, access control, surface and airborne monitoring and personal protective equipment, as appropriate, must be used.  (Refer to Section 7)

Also remember that other hazards, such as chemical and manual handling, may be present and must be controlled.

2. Pretreatment

Figure 4

Steps in pretreatment

Pretreatment of waste is the initial step in waste management after waste generation.  It consists of collection, segregation, chemical adjustment and decontamination and may also include a period of interim storage. This initial step is important as it provides the best opportunity to segregate waste streams for recycling, disposal as ordinary non-radioactive waste (when the quantities of radioactive materials are exempt from regulatory controls) or storage to allow decay (to reduce the dose rates and hence the radiological hazards during any further treatment).

Following pretreatment, the material may follow several pathways.  It may be suitable for reuse, may require treatment or may  be classified as exempt waste.

1.  Segregation

Waste segregation should occur in the workplace where the waste is generated.  Further segregation may be necessary at the waste management facility.   

Waste can be segregated according to the following characteristics.

• Active and Non-Active (see Figure 5).

• Half-life.

• Radiation types and energies.

• Appropriate treatment methods.

Segregation based on radioisotope half lives allows for the possible storage and decay of short lived radioisotopes.  When the activity is below the required clearance levels, it can be disposed of as exempt waste.  This reduces the amount of radioactive waste which requires special disposal or discharge requirements.  For example short lived radioisotopes such as Technetium-99m (6 hour half life) could be stored for a number of half-lives, after which time it could be disposed of.  Typically, ten half-lives are suggested as a start point as this gives a reduction of approximately 1/1000, however remember that the initial activity of the waste determines if ten half-lives is enough.  In the case of Technetium-99m, ten half-lives is equal to only 60 hours, or 2.5 days.  Whereas the activity of Iridium-192 (74 day half life) contaminated waste would have reduced very little in the same 2.5 day period.  Some organizations set a limit on the maximum half-live for isotopes that will be stored for decay.  As a guide 8.5 days is one figure which is used.  (This will allow for the possible storage for decay of Iodine-131.)

As well as segregating by half life, it may be necessary to segregate waste containing radionuclides that are hard to measure.  For example, low energy beta and alpha emitters would require careful monitoring using specialized techniques.

Alpha emitting waste has high radiotoxicities.  It should be segregated at the source and monitored before it joins a specialized waste stream.

Segregation of types of waste for different treatment methods, such as compaction or incineration, is also important.

Active waste binNon-active waste bin

Figure 5

Active and non-active waste bins

2.  Chemical adjustment

Chemical adjustment makes the waste safer to handle through chemical methods.  An example of this is adding a buffer to liquid waste that is too acidic for further treatment.  By adding the buffer, the pH is lowered and therefore, the waste can be moved to the next process.

3.  Decontamination

Essentially, this is checking that the outer surface of the waste container is not contaminated.  Before moving any waste container, make sure it is not contaminated by conducting a smear/wipe test on the outer surface of the container.  When the container has reached its destination, another wipe test should be carried out, particularly when transporting liquid waste.  Ordinary cleaning products can be used to decontaminate a surface.  Make sure that the paper towel/cloth used to clean a surface is disposed of/treated as solid radioactive waste.

SELF CHECK 5

Now see how much you have understood by answering the following.

1. Explain the terms disposal and predisposal.

2. What characteristics of waste are used to segregate it at the point of generation?

Now check your answers with the model answers in your workbook.

3. Treatment

Figure 6

Steps in treatment

Treatment of radioactive waste includes processes intended to improve safety or economy by changing the physical or chemical characteristics of the radioactive waste.  Basic treatment concepts and examples include:

• volume reduction  

• change of composition  

• radionuclide removal  

Treatment may generate secondary radioactive waste to be managed, e.g. contaminated filters, spent resins and sludges.

Following treatment, some material may be suitable for reuse and some may be disposed of as exempt waste.  After treatment, the waste may or may not be conditioned prior to disposal.

1.  Volume reduction

It is common practice to reduce the volume of radioactive waste to make handling and treatment (and ultimately, disposal) easier.  There are two commonly used methods, compaction and incineration.

1. Compaction

One method of minimizing the physical quantity of waste to be handled is to reduce the volume of the waste by compaction.  A high pressure press is used to ‘squash’ the waste. (See Figure 7)

Waste suitable for compaction includes paper, glass or plastic bottles, cloth, wood, metal drums.  You can probably think of a few other examples.

Compaction does not reduce the activity of the waste only its volume.

Figure 7

Drum Compactus

2. Incineration

If the waste consists of contaminated combustible materials, incinerating the materials will reduce their volume.  The incinerator must be authorized to incinerate radioactive waste.  The license should consider the dose to workers and members of the public from radioactive waste which might be discharged from the stack with hot gases.  Gas scrubbers and precipitators may be required to reduce these discharges.

The ash left from the incineration of radioactive waste will contain concentrated levels of radioactive material.  If burning a particular material reduces the volume by a factor of 1000, the concentration will now increase by a factor of 1000, assuming that none has been caught on filters or gone up the stack.  (There is less material through which the radioactive material is spread.) The ash therefore needs to be examined for disposal as solid radioactive waste.  However, if the radioactive waste is incinerated with relatively large volumes of non-radioactive waste the radioactive ash might now be diluted with the non-radioactive ash and have a lower concentration.

Incineration is used successfully for common combustible materials such as paper.  It has also been used successfully for other materials such as hydrocarbon sludge containing low levels of Radium-226 scale.

As with compaction, incineration does not reduce the activity of the waste, only its volume.  However, there is the potential for the reduction of the concentration of the ash if large volumes of non-radioactive waste are incinerated with the radioactive waste.

2. Change of composition

Chemical methods may be used to change the composition of the waste by precipitating dissolved solids from large volumes of liquid waste.  This process may also include flocculation, which causes finely divided precipitate to join together and form larger particles.   

Methods such as this are used for the treatment of large volumes of liquid waste resulting from the manufacture of radioisotopes and also in the production of rare earth materials from sands or ores containing naturally occurring materials.

A change in the composition of the waste can be followed by processes to separate the radioactive material.

3. Radionuclide removal

Activity removal is the removal of radionuclides from the waste using processes, such as sedimentation, evaporation, filtration or ion exchange.  Removing the radioactive material will usually result in volume reduction.

4. Conditioning

Figure 8

Steps in conditioning

Conditioning of waste involves processes that transform radioactive waste into a form suitable for handling, transportation, storage and disposal.  These include immobilization of radioactive waste, placing it in containers and providing additional packaging. In many cases treatment and conditioning take place in conjunction with one another.

1. Immobilization

It is important that radionuclides in stored or disposed waste are prevented from entering the environment.  Immobilization is the process in which waste is converted to a solid product that will contain the radionuclides and minimize the chance of them entering the biosphere.  The solid product is called a waste form.

Immobilization includes solidification, embedding or encapsulating in cement or bitumen and vitrification into a glass matrix or synthetic rock.

2. Packaging

In order to make the waste form safe to handle, transport or store, it may be necessary to put treated waste into some sort of container.  If one container does not provide enough protection or waste packages are to be consolidated,  a second container, called an overpack, is used.

SELF CHECK 6

Now see how much you have understood by answering the following

1. Incineration only reduces the volume of the waste, not the activity.  Why is this important to know?

2. Why would it be an advantage to incinerate combustible radioactive waste with large quantities of non-radioactive waste?

3. Match the terms change of composition, conditioning, immobilization, pretreatment, radionuclide removal, treatment, and volume reduction with the following descriptions:

1. Conversion of waste into a waste form by solidification, embedding or encapsulation.

2. Processes intended to improve safety or economy by changing the physical or chemical characteristics of waste.

3. Processes intended to transform waste into forms suitable for handling, transportation, storage and/or disposal.

4. Evaporation, filtration or ion exchange of liquid waste streams.

5. Processes that decrease the physical volume of the waste.

6. Precipitation or flocculation of dissolved solids.

7. Processes intended to prepare waste for treatment/reuse/disposal as exempt waste.

4. Does radioactive waste have to follow all of the steps in a basic waste management program?  Explain your answer.

Now check your answers with the model answers in your workbook.

20. CONTROLLING RADIOLOGICAL HAZARDS

Radiological hazards exist at all stages of waste management, from the generation of waste through to the final disposal.  You have studied methods of controlling radiological hazards in previous modules and the same techniques (e.g. time, distance, shielding, containment, etc) should be applied in handling waste.   A waste processing facility must have a radiation protection program which addresses the following issues:

• Area and personal monitoring.

• Administrative control methods.

• Physical controls.

• Record keeping and reporting.

An effective monitoring program should be set up to measure radiation and contamination (including airborne contamination) and the radiation protection program should include appropriate actions to be taken at the pre-determined monitoring levels.

21. Monitoring Levels of Radioactivity in Radioactive Waste

To ensure radioactive waste is identified and treated correctly it is important that suitable monitoring systems are in place.  These systems need to be able to identify the radioisotopes which may be in the waste and determine the levels of activity.

Monitoring is required in order to:

0. Determine if the material is considered to be radioactive waste.

1. Classify the waste and determine the disposal or discharge option.

2. Provide data to show compliance with discharge licences or authorizations.

In addition to monitoring waste, the Regulatory Authority may require the operator to collect and analyze samples from the vicinity of any radioactive discharge outlets.  To ensure that radioactivity levels in the immediate environment are acceptable.

1. Monitoring Methods

Monitoring methods will be determined by the reason for the monitoring, the physical form of the waste and the radionuclides which may be present.

Instruments used for monitoring must be able to detect the types and energies of the radiation present.

If monitoring is for the purpose of determining whether waste is exempt, then monitoring must be done very carefully and shielding within the waste must be taken into account.   

The minimum detectable level must be clearly identified and be less than the exempt level.  For example, where bags of paper waste are monitored, the following could be carried out.

0. Sealed radioactive sources of each isotope can be placed in the center of bags of waste paper similar to those that would leave a laboratory.  By taking measurements around the bag and comparing these to the source activities used, the limit of detection of the monitor for this type of waste can then be calculated.  The limit of detection can them be examined to determine if it will meet the levels you are trying to measure.

Waste that does not fit into the exempt category can be assayed using a drum scanner which uses several large scintillation detectors to count a drum of waste from all directions and can count over a long period if required.  It will be necessary to sample the waste if alpha or low energy beta emitters are present, or if the drums offer too much shielding for low energy photons to be detected.

2. Estimating The Activity Of Waste

   1. Solid waste

The activity of radionuclides in solid waste may be determined from the drum scanning method described above.  This is not always possible and it may be necessary to estimate the activity level by a process called ‘activity balance’. This is a way of determining where fractions of the total isotope activity went in the process and how much would have been discharged.

Figure 9

Activity balances

EXAMPLE

Carbon-14 compound paper waste which has been used to dry the tips of pipettes used in a radiochemical process.  The stock solution was 100 MBq of an isotope and you were able to measure some of the places where the isotope may have gone.

• 90MBq of the isotope was measured in the final product.  (Counted using a liquid scintillation counter)

• 8MBq of the isotope was measured as liquid waste (including washing the pipettes) going down the sink.

• 1.5MBq is estimated as being coated on the inside of the original stock bottle

• The process generates no volatile or gaseous materials.

Of the original 100MBq, you can account for 90 + 8 + 1.5 MBq = 99.5MBq.

Therefore, the activity you cannot account for is 100 - 99.5 = 0.5MBq.

You can assume that the paper waste contains 0.5MBq.

In some cases, the measurements made might be very limited.  Estimates of the generation of gases and contamination of items may need to be based on a knowledge of the process or from measurements taken on similar processes.

2. Gaseous and liquid waste

In situations where routine discharges are unlikely to exceed licence limits, “grab” samples can be taken of the waste stream at intervals agreed with the Regulatory Authority.  This method is dependent on a numerical estimate of the total volume of gas or liquid discharged and an assumption that the rate of discharge is constant.  The activity of the sample is measured and the value used to estimate total discharge.

EXAMPLE

A laboratory disposes of liquid radioactive waste for 12 hours a day into the site’s liquid effluent line.  The lab is required to estimate its daily discharge of radioactive materials.   

A grab sample of 0.5 L of liquid effluent is taken and measured.  The total activity of the radioisotope is 1kBq.  The liquid effluent discharge line discharges 100L h^-1 of liquid from the site.

Step 1:Calculate the concentration of the total discharge.

The concentration of the radioactive material is

Step 2:Calculate the total volume of effluent discharged per day.

12 hours per day x 100 Lh^-1 = 1200 L per day

Step 3:Calculate the activity discharged per day.

2 kBqL^-1 x 1200 L per day = 2400 kBq per day

 = 2.4 MBq per day

In some situations it may not be possible to measure the radioactive materials directly.  In those situations an estimate of the activity would need to be made.

Remember the methods of assessing the activity of materials for disposal must be approved by the Regulatory Authority.

SELF CHECK 7

Now see how much you have understood by answering the following

1. Why is it necessary to monitor radioactive waste?

2. List some important issues to consider when monitoring waste.

3. How can the activity levels of solid waste be measured?  What are some of the limitations of this method?

4. When might grab samples be an acceptable way of monitoring waste?

5. When monitoring solid waste, why might you use different types of monitors?

Now check your answers with the model answers in your workbook.

22.  Storage of Radioactive Waste

Although storage of radioactive materials is covered in Module 3.2 Transport and Storage, it is an important part of the waste management process, and some specific points must not be forgotten.

In the sequence of radioactive waste management, storage is required to provide the following;

• Operational convenience (e.g. what to do with waste in the work place)

• Storage-Storage between various stages of waste processing (interim or temporary storage).

• Safe retention of conditioned waste containing longer lived activity pending the eventual establishment of disposal facilities.

1. Operational Convenience

Waste storage facilities for unprocessed waste will depend to some extent on the types of waste, the activity level and the half-life of the radionuclides involved.  Storage in the workplace may range from simple stores with single secure cabinets to one or more dedicated rooms.  It is important that careful waste segregation and good record keeping is maintained to allow efficient movement of waste for treatment or disposal.

• Segregation of waste in the work place (point of generation) is important for dose control in waste handling.  Typically, waste is separated by type and sometimes, by isotope or half life.

Common categories of waste are dry solid waste, liquid waste, sharps (includes broken glass and metal pieces – anything that may cause a cut or puncture wound) and biohazardous waste (can include animal carcasses, animal bedding, animal waste, body fluids, tissue sample).

Decay storage is usually applied to segregated low level waste from hospitals, universities and research laboratories for common shorter lived constituents such as Technetium-99m (6 hours), Iodine-131 (8 days) Iodine-125 (60 days) and Iridium-192 (74 days).  As discussed earlier, a decay storage of ten half-lives will reduce the initial radioactivity down to 1/1000, which in many cases means below the limits for disposal, depending on the local regulatory practices.

• A planned system of collection and transfer of waste is required to prevent accumulation of waste in working areas which may lead to spread of contamination and unnecessary occupational dose exposure.

• Lined metallic or cardboard bins of less than 50 L capacity are commonly used as workplace receptacles for combustible-compactable waste.  Refuse bins with foot operated lids are particularly useful for radioisotope laboratories.

• Combustible waste should not be allowed to build up in work places.

• Bins for active waste should be clearly marked to distinguish them from bins meant for inactive waste.  

• Sharps that are used in medical procedures, radiopharmaceutical production and research using radioactive materials, such as needles, syringes, scalpels and intravenous tubing, are often considered biohazardous waste as well as radioactive.

• Metal cans are considered the best choice for metal pieces and broken glass (if sharps containers are unavailable) as they can be sealed and disposed of without further risk to those handling the waste.

• At radioisotope production facilities quantities of solid and liquid waste at intermediate activity levels can be produced.  These will require local decay storage of the short lived contaminants in carefully designed storage arrangements adjacent to or in hot cells.  Solids waste arising from production of Molybdenum -99, Iodine-125 and –131, such as contaminated spent cans and extraction columns, require some months of shielded storage prior to release for further treatment.  Liquid waste should be stored in shielded containment flasks while awaiting processing.

2. Storage

The design of a storage facility to allow the decay of short lived radioactive material or while waiting to transport the waste to a repository, will depend on the amount and activity of the waste.

Stores suitable for this type of waste can be as simple as a barriered area (provided it is secure) or a locked cupboard for small quantities.  For larger quantities of waste, or where the dose rate requires the waste to be shielded or segregated, a purpose built room or building should be used.

The general design features of these stores should comply with appropriate safety and good housekeeping requirements, including the following;

0. Containers and stores holding radioactive waste must be labeled to warn people of the presence of radioactive materials.

1. Store design should always provide security from unauthorized entry.

2. Stores for radioactive material should be away from working areas and designed so that all persons are adequately protected during both storage and transfer of material to and from the store.

3. Shielding should be provided so that radiation doses at any accessible position outside the store do not exceed the levels specified by the Regulatory Authority.  These levels are usually set at a value that ensures people in the vicinity of the store will not exceed dose limits or dose constraints.  These levels will vary depending on the location of the store and the occupancy of the area.  Shielding should take account of scattered radiation and source distribution.  Advantage should be taken of self-shielding, e.g. by arranging spent sources so that the highest activity is at the center.

4. Storage of waste should be tidy and depending on the amount of material placed in drums, racks, pallets or skids suitably planned for minimum handling.  Adequate storage capacity should be provided to allow good housekeeping and avoid waste storage in working areas.

5. Material storage must be designed to ensure smoke detection and sprinklers and drains are not obstructed.  There should be fire fighting provisions and adequate clearance below sprinkler heads.

6. Appropriate ventilation should be included in the design, particularly if waste include volatile or gaseous radionuclides e.g. Tritium, Carbon-14, Iodine-131 or gas producing nuclides such as radium or thorium.

7. Store design should take account of risks resulting from natural sources such as flood or gales.

8. Stores should take into account non-radiological hazards of the waste.  For example, the radioactive component of the waste may have decayed, but a chemical or biohazardous property may still remain.

9. Some countries do not have repositories established, so waste must be put into storage to await future disposal.  These storage facilities should include multiple engineered or natural barriers to prevent the migration of radioactive waste.  These barriers incorporate concrete or clay lined pits for the burial of drums of radioactive waste.  In addition to the normal radiological precautions such as security, contamination and dose rate controls, storing for longer time intervals has potential problems such as leaching by ground water and seismic conditions, that should be considered.  This sort of store includes the use of geological disposal sites.  These type of facilities are a specialized area and not part of this module.

23. Records

Accurate record keeping is an important part of the control of all radioactive materials.  It is important that the presence and subsequent treatment and/or disposal of radioactive waste is recorded.  This will ensure the waste can be controlled and reduces the possibility of causing uncontrolled exposure in the environment.

Records are required for the following types of waste:

• Exempt waste.

• Waste at various stages of processing.

• Waste in storage awaiting disposal.

• Liquid and airborne discharges.

• Waste sent to a disposal facility.

Each type of record should contain the following information.

0. Identification/Record number.

1. Description of waste.  

2. Nuclide identification and total activity.

3. Activity concentration (in Bq g^-1 or Bq m^-3).

4. Location (if in store).

5. Disposal route.

6. Disposal authorization.

7. Disposal date.

The format and retention of these records will be determined by the national requirements and will form a part of a quality assurance system.

SELF CHECK 8

Now see how much you have understood by answering the following.

1. What 3 reasons would there be for storing radioactive waste?

2. How should the following items be stored in the work place?

1. Paper and gloves contaminated with Co-60.

2. Syringes contaminated with I-131.

3. Liquid waste containing fission products.

3. When waste is stored to decay how many half lives are normally used as the storage time?

What fraction of the original does this number of half lives give?

4. Will this number of half lives always be enough to allow release of the waste?  Give a reason for your answer and if possible suggest an example

5. List (briefly) the requirements for a store for waste which is awaiting disposal.

6. What pieces of information would you expect to be the minimum detailed in a record of the disposal of radioactive waste?

Now check your answers with the model answers in your workbook.

KEY POINTS

• The definition of radioactive waste for regulatory purposes is material for which no further use is foreseen, that contains, or is contaminated with, radionuclide concentrations or activities greater than the clearance levels as established by a regulatory body.

• The internationally agreed objective of radioactive waste management is to deal with radioactive waste in a manner that protects human health and the environment now and in the future, without imposing undue burdens on future generations.  

• A good waste management programme minimizes the amount of radioactive waste generated, thus minimizing the hazard and maximizing the use of space in disposal facilities.

• The international classes of waste are exempt waste, low intermediate level waste and high level waste.

• Exempt waste (EW) is defined as waste that is excluded from regulatory control because its radiological hazards are negligible, i.e. its activity concentration and/or total activity are below clearance levels established by a country’s Regulatory Authority.

• Low intermediate level waste (LILW) is further divided into 2 categories - short lived waste (LILW-SL) and long lived waste (LILW-LL).   

• Short lived waste is waste containing radionuclides having a half life less than 30 years.

• Long lived waste is waste containing long lived (T_1/2 > 30 years) radionuclides having sufficient radiotoxicity in quantities and/or concentrations requiring long term isolation from the biosphere.

• High level waste (HLW) is waste that is sufficiently radioactive to require both shielding and cooling.  High level waste has concentrations of long lived radionuclides which exceed the limits for LILW-LL and the radioactive decay process produces heat at a rate above about 2kWm^-3.

• Liquid/gaseous waste may be discharged directly into the environment (typically through the water system or directly into the air), providing the levels are low.

• Solid waste is typically treated and either reused, considered exempt or disposed of accordingly.

• The IAEA plays an advisory role in radioactive waste management.

• A national framework for waste management is necessary to protect that country’s citizens (and those in surrounding countries) from unnecessary exposure to radioactive materials.  

• It is important to process radioactive waste so it can be stored or disposed of in a form that poses little or no threat to the biosphere.

• Processing waste involves pretreatment, treatment and conditioning.   

• Pretreatment is any or all of the operations prior to waste treatment, such as collection, segregation, chemical adjustment and decontamination.

• Treatment is any operation intended to benefit safety and/or economy by changing the characteristics of the waste. Three basic treatment methods are volume reduction, removal of radionuclides from the waste and change of composition.

• Conditioning is defined as the operations that produce a waste package suitable for handling, transportation, storage and/or disposal.  Conditioning may include the conversion of radioactive waste to a solid waste form (immobilization), enclosure of the radioactive waste in containers (packaging) and, if necessary, providing an overpack.

• Radiological hazards must be controlled in all stages of waste management.

• External radiation hazards are controlled by time, distance and shielding techniques.

• Containment is essential when waste material is an unsealed source.

• An effective personnel and area monitoring system must be established.

• Waste must be monitored in order to identify exempt waste, determine appropriate treatment and to demonstrate compliance with licences and authorizations.

• Instruments and monitoring methods for waste monitoring must be appropriate for the task.

• Storage at all stages of waste management must provide safety, by minimizing radiological hazards and security.

• Record keeping is an important component of waste management.

• The characteristics to be recorded for radioactive waste are Identification/Record number, description of waste, nuclide identification and total activity, activity concentration (in Bq g^-1 or Bq m^-3), location, disposal route, disposal authorization and disposal date.

APPENDIX 1

Table 6

Waste classes and recommended disposal options.

Table 7

Default values of habit and other data for external exposure, inhalation and ingestion dose estimation for a critical group in Europe.*

*Excerpt data from Table XIV, IAEA Safety Reports Series No. 19, “Generic Models for Use in Assessing the Impact of Discharges of Radioactive Substances to the Environment.”

APPENDIX 2

DISCHARGE LIMIT CALCULATIONS

In order to calculate discharge limits, it is necessary to:

• Choose a model for your situation/screening calculations.

• Consider the most significant exposure pathway.

The following examples for liquid waste and for gaseous waste show you how to calculate discharge limits in a simple way.  Remember that here we have simplified the process in order to help you understand it.

LIQUID WASTE EXAMPLE

Question:

A laboratory wants to discharge tritiated water into a local river.  The river passes a water treatment plant where the water is taken, treated (with no further dilution) and supplied to the local town as drinking water.   

The laboratory must provide information about the proposed discharge to the Regulatory Authority who will then check the details and the projected doses to the public and any potentially exposed worker, prior to issuing an authorization and setting a limit on the discharge.

The tritiated water is disposed of through a sink drain into the site effluent, which has a discharge rate to the river of 10 000 m³ per week.

Calculate the average weekly discharge activity limit.

Information:

For tritiated water, the most significant pathway into the body is via ingestion.

The default annual intake of drinking water for an adult member of the public is 600 L (or 0.6 m³).

The dose conversion factor for ingestion of tritiated water is 1.8x10^-11 Sv Bq^-1.

The river has a flow rate of 100 000 m³ per week.  (An approximate estimate could be made from the physical dimensions of the river and the linear flow rate.)

Assumptions:

Assume that the dose constraint for a member of public has been set by the Regulatory Authority as 0.1 mSv y^-1 from liquid discharges.  

Method:

Step 1:Let the maximum activity released per year be A Bq y^-1.

Step 2:Calculate the volume of water (including the discharge) passing through a point of interest per year (m³ y^-1).

The volume of water passing a given point in a year represents the effective annual dilution of the radioactive material.   

The annual dilution is:

52 weeks per year x (10 000 + 100 000) m³= 5.72 x10⁶ m³

= 5 x10⁶ m³

In order to provide a conservative estimate, the annual dilution is rounded down to 5 x10⁶ m³ instead of rounded up to 6 x10⁶ m³.    

Step 3:Find A, substituting the appropriate dose constraint.

For a dose of 0.1 mSv y^-1 (or 0.1 x 10^-3 Sv y^-1):

0.1 x 10^-3 =  

0.1 x 10^-3 = (2.16 x 10^–18 x A)

So A =  

= 4.6x 10¹³

= 46 x 10¹² Bq

= 46 TBq

Step 4:Calculate the average weekly discharge limit.

The limit for the weekly discharge for tritiated water would be:

Conclusion:

Therefore, the lab can discharge an average weekly activity of 890 GBq of tritiated water into the river without exposing an adult member of the public to an annual dose over the dose constraint imposed by the Regulatory Authority.

GASEOUS WASTE EXAMPLE

Question:

A radioisotope production plant is seeking authority to discharge gaseous Iodine-131 from its stack.   

The plant must provide information about the proposed discharge to the Regulatory Authority who will then check the details, prior to issuing an authorization and setting a limit on the discharge.

Calculate the weekly discharge activity limit for gaseous I-131.  

Information:

The stack is 60m tall and the cross-sectional area of the stack at the point of release is 0.7 m².

The velocity of the discharge is 15 ms^-1.

Let the fraction of the time the wind blows towards the point of interest be P_p.  A value of 0.25 (dimensionless) is suggested for this calculation.  This is equivalent to an occupancy factor.

To determine the critical group for an inhalation calculation, multiply the dose conversion factor for inhalation with the inhalation rate.  The critical group is indicated by the highest product.  Table 3 shows how to determine the critical group.

Table 8

Determination of Critical Group