DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 2.1

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 2

RADIATION PROTECTION

MODULE 2.1

PRINCIPLES OF RADIATION PROTECTION

CONTENTS

 OVERVIEW

This module provides information on the principles of radiation protection.  A good knowledge of the basic principles of radiation protection will help you to understand the need for an effective radiation protection program and how such a program can be implemented.

Module 1.6 explained how ionizing radiation may affect the body.  The major objective of radiation protection is to control exposures to radiation so that deterministic effects are avoided and chances of stochastic effects are minimized.  In this module you will learn how radiation protection principles can be applied in order to achieve this objective while still allowing the beneficial uses of ionizing radiation.

 

TIME ALLOCATION

It is estimated that you will need to spend five to seven hours to complete this module before commencing the final assignment.  This time estimate is for guidance only.  Do not worry if you need more time to understand the module.

MATERIALS

You will need your workbook and writing materials to take notes and answer the questions in this module.

NOW YOU ARE READY TO START WORK

 1.INTRODUCTION TO RADIATION PROTECTION

All human beings are exposed to ionizing radiation from natural and artificial sources. Exposure to natural radiation arises from both cosmic and terrestrial sources, as well as from natural radioactivity in our food and drink.  Throughout history, man has been exposed to natural radiation, and it is impossible to decide whether this radiation has been harmful or beneficial to the human species.  In contrast, artificial radiation sources have only been introduced in the last 100 years and although many benefits have been gained from their use (e.g. medical, industrial and agricultural uses), it has been realized that exposure to these sources can be harmful to us.  For this reason, a system of radiation protection has been developed to protect people from unnecessary or excessive exposure to ionizing radiation.  As the effects of ionizing radiation are better understood, this system is updated to ensure the best possible protection for both radiation workers and for members of the general public.

 2.RADIATION PROTECTION AND RISK

For the purpose of this learning material, it is important to clearly define the term radiation protection.  In simple terms, radiation protection is defined as the science and practice of limiting harm to human beings from radiation.  As we are only concerned with the effects of ionizing radiation, all references used in this module refer to ionizing radiation only.

In every activity of our daily lives humans are exposed to hazards.  For example, every time you cross the road there is a possibility that you may be hit by a car, and each time a person smokes a cigarette there is an increased chance of that person contracting lung cancer.  Although we are all aware of the possible consequences, people still cross roads and still smoke cigarettes.  This apparent disregard for our own safety may be because we have some appreciation of the probability and size of the consequence of each action; that is, we have an idea of the risk of crossing the road or smoking a cigarette.  This also applies in the field of radiation protection; in all radiological activities it is important to have some idea of the risk associated with the use of ionizing radiation.  In simple terms, risk is defined as a combination of the probability that a harmful consequence may occur and the size of that consequence.

The information used to estimate the risk of ionizing radiation is derived from detailed accident investigations and long-term epidemiological studies of populations exposed to radiation, such as the survivors of the atomic bombing of Hiroshima and Nagasaki in Japan in 1945.  Two main organizations (the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) and the Advisory Committee on the Biological Effects of Ionizing Radiation (BEIR)) are responsible for collecting and summarizing the relevant data, and for producing detailed reports on their findings.  These reports are then collated to establish risk estimates on which a system of radiological protection can be based.  The organization responsible for making recommendations based on these risk estimates is known as the International Commission on Radiation Protection (ICRP) and Figure 1 illustrates the responsibilities and relationship of these different committees.

Figure 1

The Relationship Between UNSCEAR, BEIR and the ICRP

 3.INTRODUCTION TO THE ICRP

 3.1History Of The ICRP

The International Commission on Radiological Protection (ICRP) was established in 1928.  At that time, people were concerned with the hazards associated with X-rays and radium and the ICRP was primarily responsible for protecting workers exposed to these sources.  In 1950 the ICRP’s responsibilities were extended to deal with the new problems arising from the introduction of nuclear reactors and the organization was subsequently restructured.  Since then, the interests of the ICRP have broadened to take account of the increasing uses of ionizing radiation and of practices that involve the generation of radiation and radioactive materials.  Nowadays the ICRP considers radiation protection issues which affect members of the public as well as people who work with sources of ionizing radiation.

 3.2The Role Of The ICRP

The ICRP is a non-governmental scientific organization which is responsible for making recommendations and providing guidance on the fundamental principles of radiation protection.  It brings together expertise from outside the ICRP as well as from its own committees to produce reports on the latest radiation protection methods and advances.  The ICRP intends these reports to be of help to regulatory and advisory agencies at national, regional and international levels.  Because of the differing conditions in various countries, the ICRP does not intend to provide legal texts.  Instead, the recommendations of the ICRP are designed to be incorporated into the regulations of each country as required.  In this way, the ICRP intends to provide an internationally consistent but flexible framework for a universal system of radiation protection.

 3.3Recommendations Of The ICRP

The ICRP produces regular publications concerning radiation protection issues.  In addition to reports on specific radiation protection issues such as radiation protection in medicine, dentistry and mining, more general documents are also produced.  These documents are known as Recommendations of the ICRP and they cover issues which are important to all aspects of radiation protection.  The first Recommendations of the ICRP appeared as Publication 1 in 1959 and many other publications have been produced since then.  Each publication represents an increase in knowledge of the risk associated with radiation exposure.  The Recommendations are reviewed annually, and amendments and clarifications are issued as required.  Figure 2 shows the sequence and dates of the Recommendations of the ICRP along with their publication numbers.

Figure 2

The Sequence of The Recommendations of The ICRP

The most recent publication was produced in 1990 and is commonly known as ICRP 60.  The Recommendations of this document are discussed later in Section 8.

 4.BASIC SAFETY STANDARDS

In 1990, the Inter-Agency Committee on Radiation Safety (IACRS) was established to further promote consistency in the practical application of radiation protection.  This committee consisted of a group of international organizations with a common interest in radiation protection. As a result, a detailed document (called the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources) was prepared.  This document, commonly known as the Basic Safety Standards (BSS), is intended to serve as a practical guide for the design and implementation of an effective radiation protection program.  It contains standards which are primarily based on the principles of radiation protection recommended by the ICRP.  It also takes into account nuclear safety principles as recommended by the International Nuclear Safety Advisory Group (INSAG), a group which has been producing Nuclear Safety Standards (NUSS) specifically for nuclear installations since 1985.  Many of the principles recommended by INSAG are relevant to other radiation sources and installations.  Figure 3 illustrates the development of the Basic Safety Standards.

Figure 3

The Development of the Basic Safety Standards

Like the ICRP recommendations, the Basic Safety Standards are intended for guidance only and not as legal text.  The standards are designed such that they may be incorporated into the legislation of individual countries as required to form a practical basis for the specific radiation protection needs of that country.

SELF CHECK 1

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

1. Match the terms radiation protection, risk, and ICRP with the following descriptions:

 a)  The organization which makes recommendations and provides guidance on the fundamental principles of radiation protection.

 b)  The probability of harm or damage.

 c)  The science and practice of limiting harm to human beings from radiation.

2. On which evidence are risk estimates based?

3. How often does the ICRP review its recommendations?

4. Which ICRP Publication contains the most recent Recommendations of the ICRP?

5. What is the function of the Basic Safety Standards?

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

 5.PRACTICES AND INTERVENTIONS

Human activities which increase the overall exposure to radiation are called practices.  Practices may either introduce a new facility or procedure (e.g. a new nuclear medicine procedure, a new chest X-ray screening procedure or new mineral sands processing plant) or change an existing one in such a way that more radiation is produced or more radioactive waste is released to the environment (e.g. increasing the capacity of an existing nuclear power plant or changing a chemical process such that an increase in volatile radioactive materials released to the air is observed).

Interventions are any human activities that are intended to reduce the existing radiation exposure, or the possibility of exposure.  These apply to both chronic exposure situations such as radon in buildings, and emrgency situations.

 6.EXPOSURES

 6.1Introduction To Exposures

Exposure is defined as the actual process of a human being exposed to radiation or radioactive materials.  Exposure to ionizing radiation will lead to a radiation dose and the amount of dose received depends on many factors.  You will learn more about this in Module 2.5 Personal Dosimetry.

Exposure can be either external (i.e. exposure from a source outside the body) or internal (i.e. exposure from a source inside the body).

 6.2Potential Exposures

The idea of potential exposure was introduced to cover the possibility, but not certainty, of exposure.  For example, transport accidents are possible, but not certain to occur.  In the routine use of radioactive materials, accidents can always happen but they rarely occur.  A radiation protection system must try to control the probability and the size of the potential exposure.

 6.3Types Of Exposure

The ICRP refers to three types of exposure:

1. Occupational exposure is the exposure of a person in the workplace and mainly as a result of the work they perform.  It can be controlled at three levels; at the source (e.g. shielding); in the environment (e.g. ventilation); and at the level of the worker (e.g. work practices, protective clothing).

2. Medical exposure is the exposure of a person as part of a medical diagnosis or treatment.  The exposure is controlled at the same three levels as for occupational exposure but is included as part of the diagnosis or treatment plan, not as part of a radiation protection system.

3. Public exposure is the exposure of a person by means other than occupational or medical exposure.  The controls should be applied at the source. They should only be applied to the environment or to the individuals if source controls are not effective, as, for example, in a serious accident such as Chernobyl.

 6.4Levels Of Exposure

In order to compare the amount (or level) of exposure from different sources, we need to be able to quantify exposure.  For the purpose of this module, we use the level of dose received from each source (in units of millisieverts) to provide a comparison.  The units and definitions of dose will be further discussed in Module 2.5 Personal Dosimetry.

Figure 4 compares the average levels of occupational and medical exposures with the average radiation levels from natural sources.

* From UNSCEAR Publication 1993

Figure 4

Comparison of Levels of Exposure

 7.DOSE LIMITS AND DOSE CONSTRAINTS

 7.1Dose Limits

Dose limits are defined as the level of dose (i.e. the amount of radiation received or absorbed) which must not be exceeded under normal controlled conditions.  Occupational dose limits were originally introduced in 1934 to avoid directly observable non-malignant effects such as changes to the skin.  By 1950, it was recognized that cancer, particularly leukaemia, was one of the risks of exposure to ionizing radiation.  Consequently, the set dose limits were intended to reduce the incidence of cancer and hereditary effects, and to prevent deterministic effects.  Over the years, these limits have been expressed in a variety of ways; so comparisons are not easy.  In broad terms, however, the annual occupational exposure of the whole body was reduced by a factor of about three between 1934 and 1950, and by a further factor of three, to the equivalent of 50 millisieverts per year, in 1958.  Since then, new information on the risk of radiation-induced cancers has become available and occupational exposures were reduced by a further factor of 2.5 in 1990 to the present limit.

Limits on public exposure were first introduced by the ICRP in 1959.  Since then, these limits have been reduced in line with occupational dose limits to the current limit of 1 millisievert per year.  Dose limits recommended by ICRP Publication 60 are discussed in Section 8.3 of this module.

 7.2Dose Constraints

People may be exposed to radiation from several different sources or from a source which may increase with time (e.g. a radioactive store or waste repository).  In this case it is necessary to restrict the exposure from each source to ensure that regulatory dose limits are not exceeded.  A dose constraint is the upper restriction on dose that may be received from a single source, and these constraints are particularly applicable to public exposure.  Statutory authorities control potential doses to the local population by applying suitable dose constraints.  For example, a member of the public may live close to a radiotherapy practice and a nuclear power plant.  The statutory authority will decide that the radiotherapy practice can only give rise to a certain percentage of the 1 millisievert public dose limit, and the nuclear power plant may give rise to another part of the 1 millisievert.  This concept of dividing the dose amongst the sources of ionizing radiation is referred to as applying dose constraints.

SELF-CHECK 2

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

1. Match the terms practices, interventions and potential exposure with the following descriptions:

 a)  Human activities which decrease the overall exposure to ionizing radiation.

 b)  The possibility of exposure to ionizing radiation.

 c)  Human activities which increase the overall exposure to ionizing radiation.

2. Give two examples of practices and two examples of interventions.

3. What are the three types of exposure referred to by the ICRP and when do they apply?

4. Define the terms dose limit and dose constraint.

5. Compare the terms dose limit and dose constraint by giving examples of how they are used.

6. Using the information in Section 6.4, compare:

 a)  the average annual exposure of workers in the nuclear industry with the average annual exposure from natural background radiation;

 b)  the maximum annual exposure of members of the public from non-natural sources with the average annual exposure from natural background radiation; and

 c)  the average annual medical exposure from diagnostic procedures with the average annual exposure from natural background radiation.

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

 8.THE SYSTEM OF RADIATION PROTECTION

The system of radiation protection recommended by the ICRP in Publication 60 is based on three major principles.  These principles are known as justification, optimization, and dose limitation.  The ICRP recommends that these three general principles are applied in the control of new and current practices.  Figure 5 illustrates the relationship between the radiation protection system and the three principles.  Each principle is discussed further in Sections 8.1, 8.2 and 8.3.

 Benefit > Risk     ALARA          Numerical Dose         Limits

Figure 5

The Three Major  Principles of Radiation Protection

It is important to remember that none of the principles should be used on their own.  An effective radiological protection system should use the three principles to ensure that all radiation doses are kept as low as possible.

 8.1Justification of a Practice

In terms of radiation protection, decisions concerning the adoption of new practices or the continuation of existing practices must take into consideration the risks of radiation effects as a result of that practice.  A practice which involves exposures or potential exposures should only be adopted if it is likely to produce sufficient benefit to the individual or to society to outweigh the detriment or harm to health it may cause.  For example, a mass screening program for the detection of breast cancer must be shown to have sufficient benefit by detecting the cancer early to outweigh the additional risk associated with the increased radiation doses.  Figure 7 illustrates the principle of justification of a practice.

Figure 6

The Principle of Justification of a Practice

 8.2The Optimization of Protection

Individual doses, the number of people exposed and the likelihood and magnitude of potential exposures should all be kept as low as reasonably achievable, economic and social factors being taken into account.  This is known as the ALARA principle.  It means that radiation sources and installations should be provided with the best practical protection and safety measures possible, with all social and economic factors being considered.  Protection and safety should be optimized.  For example, before additional shielding is added to a nuclear installation, it should be shown that individual doses and the number of people exposed will be decreased sufficiently to justify the potential doses to the people installing the shielding and the cost.  It is important to remember the principle of ALARA as shown in Figure 7.

Figure 7

The Principle of ALARA

 8.3Individual Dose Limits

The ICRP recommends that the exposure of individuals should be subject to dose limits.  The occupational dose limits are aimed at ensuring that no individual is exposed to unacceptable risks as well as preventing any deterministic effects and minimising the chance of stochastic effects.  Figure 8 shows the principle of occupational dose limitation as adopted by the ICRP.

Figure 8

The Principle of Occupational Dose Limitation

For members of the public, the ICRP recommends a dose limit which is approximately the same as the average dose from natural sources, excluding radon, at sea level.

Occupational dose limits and public dose limits apply to exposures resulting from practices, and exclude medical exposures and natural background.The annual occupational dose limit for effective dose is 20 mSv with a further provision that it should not exceed 50 mSv in any single year.  The annual public dose limit is 1 mSv, but in special circumstances a higher public dose could be allowed as long as the average over five years does not exceed 1 mSv per year.

The term effective dose applies to a dose to the whole body.  There are additional dose limits for the lens of the eye and the skin, 150 mSv and 500 mSv respectively, which provide protection against deterministic effects, such as cataracts in the lens of the eye, erythema (skin burns) and desquamation (shedding of layers of skin).  The effective dose limit provides sufficient protection against stochastic effects in these organs.  The 500 mSv skin dose limit is averaged over any one square centimetre, regardless of the area exposed.

The current dose limits recommended in ICRP Publication 60 are summarized in Table 1.  Doses are given in millisieverts (mSv) but do not worry about the terms and units.  These will be explained in Module 2.5, Personal Dosimetry.

Table 1

Dose Limits Recommended in ICRP Publication 60

 8.3.1The Occupational Exposure Of Pregnant Women

It is considered by the ICRP that, if the recommended system of protection is in place, then there is no need to have special occupational dose limits for women who may become pregnant.  However the foetus should be treated like a member of the public and it may be necessary to adapt the woman’s working conditions if she becomes pregnant to ensure that this occurs.  In particular, the type of work the woman does should not carry a significant probability of high accidental exposure or intake of radionuclides.

The ICRP recommends that once pregnancy has been declared, the equivalent dose to the surface of the woman’s abdomen should be limited to 2 mSv and the intake of radionuclides should be limited to about one twentieth of the intake limit of other radiation workers.  You will learn about restricting the amount of radionuclides entering the body in Module 2.3 Protection from Internal Hazards.

SELF-CHECK 3

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

1. What are the three basic principles of radiation protection upon which a system of radiation protection should be based?

1. How may a practice be justified?

1. What is the ALARA principle?

1. What are the three criteria for setting dose limits?

1. What types of exposure are not included in the occupational and public dose limits?

1. What is the annual effective dose limit for the following people?

a)  A person who works with radioactive materials.

 b)  A person who lives next to a nuclear power plant but who does not work there.

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

RESEARCH ASSIGNMENT

With the help of your supervisor, find out what the dose limits are for your country.  Write your answers in the space provided in your workbook.

 9.IMPLEMENTING A SYSTEM OF RADIATION PROTECTION

The most important principle of radiation protection is to keep all doses as low as possible while still allowing the beneficial use of ionizing radiation.  The recommendations of the ICRP can be applied at several levels in order to control the hazards from radiation.  These levels are as follows:

1. Regulatory.

2. Management.

3. Operational.

 9.1Regulatory Requirements

The formulation and implementation of regulations and legislation varies from country to country.  Regulations provide a link between the ICRP recommendations and their implementation in the workplace.  The country’s regulators (an agency or government department) should be responsible for assessing whether a practice is justified, and for prohibiting it if there is no justification.  The regulations should set a satisfactory standard of protection which can be applied to all justified practices.

 9.2Management Requirements

It is not possible for workers to implement satisfactory standards of safety without the support and direct involvement of management.  Hence, a formal management structure should be introduced to deal specifically with the issues of radiation protection.  One of main responsibilities of this management is to encourage a good attitude to safety and a recognition that safety is a personal responsibility.  In addition, the management should optimize protection by defining responsibilities clearly and providing clear and simple operating instructions.  It should also take potential exposures into account and provide a system for safety analysis to identify possible causes of accidents and limit the probability and effect of such accidents.

 9.3Operational Requirements

In addition to regulatory and management requirements, control is necessary at the operational level to ensure safe practices in the workplace.  These practical considerations should include such aspects as –

4. safe storage of radioactive materials;

5. safe transport of radioactive materials;

6. use of time, distance and shielding to reduce radiation fields;

7. use of containment to limit the spread of radioactive materials into workplaces and the public environment;

8. adequate maintenance of plant and equipment to reduce the probability of failure; and

9. safe management of radioactive wastes.

The operational requirements listed above will be covered in detail in later modules.  In addition there is a requirement on individual workers that they accept responsibility for their own safety.  They must use all protective equipment supplied and work according to safety procedures.

SELF-CHECK 4

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

1. What are the three levels at which radiation protection principles can be applied?

2. What is the purpose of regulations?

3. Name two responsibilities of the management structure.

4. Name three practical requirements at the operational level.

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

KEY POINTS

10. Radiation protection is defined as the science and practice of limiting harm to human beings from radiation.

• Risk is defined as a combination of the probability that a harmful consequence may occur and the size of that consequence.

11. The risk from ionizing radiation is determined by the study of populations exposed to high doses.

12. UNSCEAR and BEIR publish information on the risk of radiation-induced cancer in humans.

13. The ICRP prepares reports on specialised areas of radiation protection and produces general recommendations for the establishment of radiation protection systems.

14. The Basic Safety Standards are practical working documents for the establishment of a radiation safety system.

15. Human activities which increase the overall exposure to radiation are called practices.

16. Human activities which decrease the overall exposure are called interventions.

• Exposure is defined as the actual process of a human being exposed to radiation or radioactive materials.

• Potential exposure is used to describe the possibility, but not certainty, of exposure.

4. Occupational exposure is the exposure of a person in the workplace and mainly as a result of the work they perform.

5. Medical exposure is the exposure of a person as part of a medical diagnosis or treatment.

6. Public exposure is the exposure of a person by means other than occupational or medical exposure.

7. Natural background exposure is the exposure from natural sources.

17. Dose limits are defined as the level of dose which must not be exceeded under normal controlled conditions.

18. Dose constraints are a means of controlling doses from several sources.

19. The three basic principles of radiation protection are justification, optimization and dose limitation.

20. For justification of a practice, the benefits must be shown to outweigh the risks.

21. Optimization of protection requires individual doses, the number of people exposed and the likelihood and magnitude of potential exposures to be kept as low as reasonably achievable, economic and social factors being taken into account.  This is known as the ALARA principle.

22. Dose limits are necessary to ensure that no individual is exposed to unacceptable risks.  They are set such that deterministic effects are prevented and the chance of stochastic effects is minimized.

• It is considered by the ICRP that the foetus of a pregnant woman should be treated as if it were a member of the public.

• The ICRP recommends that once pregnancy has been declared, the equivalent dose to the surface of the woman’s abdomen should be limited to 2 mSv and the intake of radionuclides should be limited to about one twentieth of the intake limit of other radiation workers.

23. The principles of radiation protection should be implemented at three levels: regulatory, management and operational.

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 2.2

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 2

RADIATION PROTECTION

MODULE 2.2

PROTECTION FROM EXTERNAL RADIATION HAZARDS

CONTENTS

 OVERVIEW

Radiological hazards are divided into external and internal radiation hazards and both of these can lead to exposure.  External radiation hazards are due to radioactive materials outside the body.  Internal radiation hazards exist when there is the potential for radioactive materials to enter the body.  The radiation protection methods for each type of hazard are different and this module covers the protection from external radiation hazards only.  The next module, Module 3.3, Protection from Internal Radiation Hazards, will deal with the protection methods for internal radiation hazards.

In this module, you will learn about the three main methods of protection against external radiation hazards – those of time, distance and shielding.  You will also learn how to apply these methods for protection against the common types of radiation emitted by sealed sources and by radiation apparatus.

 

TIME ALLOCATION

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

MATERIALS

You will need to have writing materials to take notes and answer questions and assignments in the module.  A scientific calculator would assist with the calculations.

NOW YOU ARE READY TO START WORK

 10.EXTERNAL RADIATION HAZARDS

 10.1Introduction

External radiation is radiation from a source outside the body.  An external radiation hazard exists when ionizing radiation from a source outside the body has the potential to cause harm.  This is different from an internal radiation hazard which exists when there is a potential for exposure from radioactive materials entering the body (see Module 2.3 Protection from Internal Hazards).  We use different methods to control the external and internal radiation hazards.

 10.2The Effect of Radiation Type on External Radiation Hazard

Alpha and beta particles, gamma rays, x-rays and neutrons are all types of ionizing radiation (see Module 1.4 Interaction of Radiation With Matter).  However, not all of these are external radiation hazards.  The degree to which each is an external radiation hazard is discussed below.

Alpha particles have only a very short range in air (a few centimetres), and are not considered an external radiation hazard since they cannot penetrate the outer layers of the skin.

Beta particles are more penetrating than alpha particles, with the penetrating power of beta particles being dependent on their energy.  Energetic beta particles can travel several metres in air and can also penetrate the outer layers of the skin to a few millimetres (e.g. 1 MeV beta particles close to the skin can penetrate approximately 5 mm).  If a solution containing a beta emitting radionuclide is spilled onto the skin, even a small drop may give a large dose to the skin tissue.  The only other organ of importance in the case of external beta irradiation of the eye – beta particles are not considered to be an external radiation hazard to any other body organs.  Bremsstrahlung radiation arising from high energy beta interactions with high atomic number materials will lead to the production of x-rays which are a greater external radiation hazard than the original beta particles.

X- and gammarays are forms of short wavelength electromagnetic radiation which will penetrate all organs of the body and are a significant external radiation hazard.  The energy of the x-ray and gamma photons is an important factor in determining the magnitude of the external radiation hazard.

Neutrons are also highly penetrating.  They give energy to the body as they are scattered in body tissues.  Neutrons are an important external radiation hazard which require careful control.

Table 1 summarises the relative hazard of the different types of radiation from sources outside the human body.

Table 2

Relative External Radiation Hazards

Radiation Type	Relative External Radiation Hazard
Alpha Particles	None
Beta Particles	Slight
Gamma Rays	Severe
X-Rays	Severe
Neutrons	Severe

 10.3Sources of External Radiation Hazards

External radiation hazards can be produced in two ways.  One way is from equipment or devices, that when operated, produce ionizing radiation.  The second way is from radioactive materials.

X-ray generators are a common ionizing radiation emitting device.  When they are turned on, x-rays are produced and the device represents an external radiation hazard.  However, when turned off, the production of x-rays is stopped and the external radiation hazard is eliminated.

In contrast, beta particles, X- and gamma rays emitted from radioactive materials represent a continuous external radiation hazard.  Radioactive materials cannot be turned off but they can be placed in containers or surrounded in a shielding material that will reduce the external radiation hazard to an acceptable level.

SELF-CHECK 1

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

1. What is the difference between an external and internal radiation hazard?

1. Why are alpha particles not considered to be an external radiation hazard?

2. For which organs can beta particles be an external radiation hazard?

3. Which types of radiation can penetrate the body?

4. What is the major difference between X-ray generators and radioactive materials as external radiation hazards?

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

 11.CONTROLLING THE EXTERNAL RADIATION HAZARD

 11.1Basic Techniques

There are three basic methods for controlling external radiation hazards.  These are:

1. Time.

2. Distance.

3. Shielding.

Application of these methods can reduce the dose received as a result of exposure to external ionizing radiation.  Each method is discussed in detail in the following sections.

 11.1.1Time

Controlling time is an important method for reducing the exposure to ionizing radiation.  By reducing the time spent working with or being exposed to radioactive materials, the dose received can be minimized.  In simple terms, the radiation dose received by a person working in an area with a certain dose rate is dependent on the time spent in that area.

This relationship is given by Equation 1:

D = R x T[1]

where   D = Dose received

R = Dose rate

T = Time of exposure

Hence, if the time of exposure to a fixed dose rate source is decreased, then the overall dose received is also decreased. This is illustrated by Example 1.

Example 1

Question

An occupationally exposed person usually spends 3 hours working on a process which has an accompanying dose rate of 5 Sv h^-1 (i.e. 5 microsieverts per hour).  The process is improved such that the process time is reduced to 1.5 hours.  What is the difference in dose received by the worker?

Answer

Using Equation 1:

Before the change in process, the total dose received was:

D = 5 Sv h^-1 x 3 h = 15 Sv

After the change in process, the total dose received is:

D = 5 Sv h^-1 x 1.5 h = 7.5 Sv

Hence, the difference in dose received is:

15 Sv - 7.5 Sv = 7.5 Sv

It can also be seen from Example 1 that, for a fixed dose rate:

If you halve the exposure time, the dose received is halved.

Conversely:

If you double the exposure time, the dose received is doubled.

It is important to remember these rules both for your own protection and for the protection of other radiation workers.

The relationship in Equation 1 can also be used to ensure that dose limits and constraints are not exceeded in the workplace.  Controlling the hourly dose rates of occupational workers, ensures that annual limits or constraints are not exceeded.  This is illustrated in Example 2.

Example 2

Question

If a radiation worker spends 40 hours per week, 50 weeks per year working with sources of radiation and the annual dose constraint is 10 mSv, what is the maximum hourly dose rate allowed?

Answer

Maximum dose allowable = 10 000 Sv

Time of Exposure = 40 x 50 = 2 000 hours

From Equation 1:

10 000 Sv = R x 2 000 h

Hence

 R = 10 000 Sv = 5 Sv h^-1

2 000 h

The maximum dose rate allowed is therefore 5 Sv h^-1.  To ensure that workers do not exceed the annual dose constraint, their exposure should be controlled to an average of 5 Sv h^-1.

The concept of time is also useful in planning work that may lead to external radiation exposure.  Often it may be useful to practice a technique without an actual radiation source being present.  This should help to reduce the exposure time and hence the total dose received.

 11.1.2Distance

Controlling distance is another useful method of external radiation control.  In simple terms, the greater the distance from a source of penetrating radiation, the less the overall exposure.

The relationship between the dose rate from a point source (a source which is small in size) and the distance from the source is given by Equation 2:

R = k[2]

d²

where   R = Dose rate

d =  Distance from the source

k =  A constant value for a particular radioactive source.

The relationship in Equation 2 is known as the inverse square law and you will gain practical experience of this law in Practical Activity 1 of this module.

Equation 2 can be rewritten to form Equation 3:

R d² = k[3]

As k is a constant value for a particular source, if the source is not changed then we can write Equation 4:

R₁ d₁² = R₂ d₂²[4]

where   R₁ = Dose rate at distance d₁ from the point source

R₂ = Dose rate at distance d₂  from the point source

In practice, we can assume that physically small sources are, in effect, point sources provided that the distance from the source is at least ten times the largest dimension of the source.

In radiation protection, distance is often used to reduce exposure to ionizing radiation, e.g. by restricting access close to sources or by using long handling tools (e.g. tongs).

If we know the dose rate at a particular distance from the source, then it is possible to calculate the distance at which dose rates are considered acceptable.  This concept is illustrated in Example 3.

Example 3

Question

The dose rate at 2 m from a gamma source is 125 Sv h^-1.  At what distance will it give an acceptable dose rate of 5 Sv h^-1?

Answer

Using Equation 4 and substituting the values gives:

125 x 2²  = 5 x d₂²

Rearranging this equation gives:

 d₂²  = 500 = 100

5

d₂  = 100 m = 10 m

Hence, by increasing the distance from the source from 2 m to 10 m, we can reduce the dose rate considerably.

A useful way of remembering the distance principle is:

When the distance from a radiation source is doubled, the dose rate is reduced to a quarter of its original value.

Again, it is important to remember this rule in order to reduce exposure to yourself and your colleagues.

PRACTICAL ACTIVITY 1

Gamma Radiation and the Inverse Square Law

In this practical activity, you will measure the dose rate at various distances from a gamma source in order to investigate the inverse square law.  Details of the procedure to follow are given in your workbook.  You should contact your supervisor to arrange a suitable time to do the practical.

 11.1.3Shielding

Example 3 above gives a good illustration of how distance may be used to reduce the dose rates and hence overall dose received.  However, in many situations it is not possible to work with a source if it is more than a metre away from you.  A more practical method for reducing exposure in workplace situations is to shield the source.  By using this method, dose rates may be reduced whilst still allowing work to be performed.

The amount and type of shielding material required depend on the following factors:

1. The type and energy of radiation.

2. The activity of the source (or intensity of radiation from a generator).

3. The acceptable dose rate outside the shielding.

In Section 1.2 you learnt about the penetrating power of different types of radiation.  This is summarized in Figure 1.

Figure 9

The Penetrating Properties of Ionizing Radiation

Alpha particles have a short range in air (a few centimetres) and are easily absorbed so they do not require shielding.

Low energy beta particles such as those from tritium (maximum energy 0.019 MeV) are also easily absorbed and do not require shielding.

However, higher energy beta particles such as those from phosphorus-32 (maximum energy 1.7 MeV) are more penetrating and interact with shielding materials to produce bremsstrahlung radiation.  Beta shields constructed of a low atomic number material (e.g. perspex) will reduce the intensity of beta particles and minimize the production of bremsstrahlung radiation.  It may, however, still be necessary to use additional shielding to protect the worker from the bremsstrahlung.  In which case, perspex surrounded by lead is recommended.

It is important not to underestimate the hazard of unshielded beta sources as they are capable of imparting large doses (especially to the hands and fingers) if they are handled directly.  This is particularly important for people handling vials or syringes containing beta emitters in solution.  For example, a small volume of solution containing a high level of activity of phosphorus-32 may give dose rates at the surface of thin-walled plastic tubes in excess of 1 mSv min^-1.

X- and gamma rays are much more penetrating than either alpha or beta particles.  Shielding with materials of a high atomic number (e.g. concrete, lead or iron) is necessary to attenuate the radiation and reduce the dose rate to acceptable levels.  The amount of attenuation is dependent on the shielding material, its thickness and the energy of the incident photons.

Shielding for neutrons is more complex due to the wide range of energies and the different reactions that may occur.  The process of shielding neutrons involves firstly the slowing of energetic neutrons by elastic scattering, and then their capture.  For these two actions to occur, a suitable shielding material should incorporate a combination of material with a high hydrogen content (e.g. water, paraffin wax, polythene, concrete) to slow down the neutrons plus boron to absorb them.  Borated paraffin wax is the name given to a mixture of paraffin wax and boron which may be used as a small area shield.   The capture reaction with boron-10 is written as follows:

¹⁰  B (n,) ⁷ Li

This means that the boron-10 nucleus absorbs a neutron, emits an alpha particle and becomes a lithium-7 nucleus.  The alpha particle is easily absorbed in the surrounding material.  The reaction also produces a gamma ray with an energy of 0.48 MeV which may require additional shielding.  Other neutron capture reactions can produce high energy gamma radiation.  An example of such a reaction is neutron absorption by iron-58.  This is written as follows:

⁵⁸ Fe (n, )  ⁵⁹ Fe

This reaction produces a gamma ray with a high energy of 7.6 MeV.  Hence, iron is not a good shield in a nuclear reactor environment.

Table 2 shows shielding materials which are recommended for different types of radiation.  The methods used for estimating the thickness of shielding required for beta particles, x- and gamma rays will be discussed in Section 3 of this module.

Table 3

Recommended Shielding Material

Radiation Type	Recommended Shielding  Material
Alpha Particles	None
Low Energy Beta Particles	None
High Energy Beta Particles	Perspex surrounded by lead
X- and Gamma Rays	Concrete, lead, iron
Neutrons	Concrete, water, polythene, borated paraffin wax

Note that the choice of shielding material is often related to cost, space and convenience.  Lead is not easy to work with, is poisonous and, in large sheets, can sag without good support.  On the other hand, concrete is cheaper and easier to handle but large thicknesses may be required for effective shielding.

SELF-CHECK 2

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

1. What are the three basic methods that can be used to control the external radiation hazard?

2. An occupationally exposed person usually spends 15 hours working on a process which has an accompanying dose rate of 5 Sv h^-1 (i.e. 5 microsieverts per hour).  The process is improved such that the process time is reduced to 7.5 hours.  What is the difference in dose received by the worker?

3. If an occupationally exposed worker spends 20 hours per week working with radioactive materials and the maximum acceptable external dose for that person is 200 Sv per week, what is the maximum hourly dose rate allowable?

4. The dose rate at 1m from a certain gamma source is 900 Sv h^-1.  At what distance will the dose rate be 25 Sv h^-1?

5. What factors are important when considering the amount and type of shielding materials required for external radiation?

6. Give a suitable shielding material for each of the following radiations:

1. Alpha particles.

2. Low energy beta particles (e.g. tritium).

3. High energy beta particles (e.g. phosphorus-32).

4. X-rays and gamma rays.

5. Neutrons.

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

 11.2Administrative Controls

An administrative control is a method of administration which prevents or minimizes exposure to the external radiation hazard.  Administrative controls include the following procedures:

1. Classification of areas.

2. Display of clear signs for each classified area.

3. Radiological protection training for workers and managers.

4. Workprocedures which incorporate the use of time, distance and shielding.

5. Local rules (e.g. restriction of access to certain areas) and conditions of work (e.g. requirement to wear an alarming dosimeter).

6. Maintenance of source inventories for each area.

7. Radiation safetyauditing system which includes safety assessment of work procedures, plant and equipment.

8. Use of investigation levels for personal dose monitoring and workplace monitoring results.

The controls of time, distance and shielding have already been discussed in Section 2.1.  The classification of areas, the international radiation warning symbol and the use of investigation levels will be discussed in the next three sections.  The other methods of administrative controls will be discussed in the relevant modules of Parts 4 and 5 of this distance learning material.

 11.2.1Classification of areas

Work areas can be classified according to the level of the external radiation hazard.  (They may also be classified according to the level of the internal hazard.  This will be discussed in Module 2.3 Protection from Internal Hazards.)  A classification system in which areas may be classified as controlled or supervised is recommended in the Basic Safety Standards.

A controlled area is one where specific protection measures and safety provisions are or could be required to control normal exposures and to prevent or limit potential exposures.

A supervised area is where occupational exposure conditions are kept under review but no specific protection measures are required.

An unclassified area is where no protection methods are required and the occupational exposures do not need to be kept under review.

Generally, the classification of areas varies in accordance with the particular needs of the workplace and is usually based on operational experience and judgement.  One way of classifying areas is with respect to the potential annual dose to workers.  Table 3 shows an example of how potential annual doses can be used in the classification of areas.

Table 4

Example of Area Classification with Respect to Annual Dose

In practice, areas are usually classified according to the measured dose rate at the boundary of the area.  Dose rate values are determined by considering the potential dose which is acceptable as an annual dose together with the amount of time spent by workers in the classified area (occupancy time).  The occupancy time is often written as a fraction of the total work time and is known as the occupancy factor.  For example an occupancy factor of 0.2 means that a worker spends one fifth (or 20%) of their working time in a particular area.  This corresponds to 400 working hours during one year (assuming a standard working year of 50 weeks, 40 hours per week).  Table 4 shows how areas could be classified according to dose rate if an occupancy factor of 0.2 is assumed.

Table 5

Example of Area Classification with Respect to Annual Dose Rate

 *  Assuming an occupancy factor of 0.2

The dose rate values given in Table 4 have been calculated by dividing the annual dose (in microsieverts) given in Table 3 by 400 hours.

Note that the values given above are an example only and the dose rate approach to area classification must be used very carefully, taking into account the individual workplace conditions.  It is very difficult to assess an average dose rate, and information about the length of time the dose rate stays at or above the defined levels along with the risk of potential exposure should be considered.  However, this method, if applied carefully, does provide adequate control of the external radiation hazard.

Once areas have been classified in a workplace, these areas should be clearly marked with visual signs to ensure that workers and visitors are aware of the classification in that area and what the classification means .  In addition, the boundaries to each area should be clearly defined.

It should also be noted that when working in classified areas with high dose rates, additional administrative controls may need to be applied.  For example, local rules may include a written system of work which limits the time spent in the area on a particular type of work.

 11.2.2The radiation warning symbol

As human senses cannot detect ionizing radiation, it is important to use warning signs wherever an external radiation hazard is present.  Warning signs should be placed on the source itself as well as in the vicinity of the source and at barriers/boundaries to the area where the source is stored.  The international symbol for ionizing radiation is the trefoil as shown in Figure 2.

Figure 10

The Radiation Warning Symbol

This symbol can be used to warn of radiological hazards in areas or on sources themselves.  The trefoil is internationally accepted as black on a yellow background.  However, these colours may vary in some countries.

 11.2.3Investigation levels

As part of an effective radiation protection program, routine monitoring of the workplace and personnel should be performed.  The results of these measurements should be reviewed regularly and any irregularities or differences from normal measurement values should be noted.  In addition, investigation levels should be set such that for any measured value at or above a certain level, an investigation into the cause takes place.  For example, a routine area survey indicates an unusually high dose rate of 50 Sv h^-1 at the surface of a working facility.  The usual dose rate is less than 5 Sv h^-1 and the investigation level is set at 10 Sv h^-1.  In this case, an investigation into the cause of the high dose rate should be performed.  This investigation should include such things as an examination of work practices, equipment and/or plant to identify the cause.  Once the reason for the high dose rate has been established, preventative measures should be taken to ensure that the potential exposure of workers remains as low as reasonably achievable at all times.  In this way, the exposure to external radiation hazards can be controlled on a routine basis.

In practice, investigation levels are usually defined during the planning of a radiation protection program.  They should however be reviewed at regular intervals to ensure compliance with relevant dose limits and constraints and to ensure that exposures are kept as low as reasonably achievable.

 11.3Physical Controls

Although administrative controls provide a basis for a radiation protection program, these controls rely on familiarity with written procedures.  If a new worker is not aware of these procedures, or these procedures are not followed correctly for any other reason, then the potential exists for accidental exposure to ionizing radiation.  In order to ensure that the chance of accidental exposure is as low as possible, physical controls are employed in the workplace.  These controls are essentially physical barriers which incorporate a range of radiation protection techniques.  The following list gives some useful examples of physical controls:

4. The use of engineering interlocks which restrict or prevent access to hazardous area.

5. The incorporation of fixed shielding into the design of plant and equipment.

6. The use of remote manipulators to eliminate direct handling and provide increased source – operator distance.

7. The use of preset timers in the case of radiography equipment to control exposure times.

 SELF CHECK 3

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

1. Draw the international radiation warning sign, indicating its colour and background colour.

2. Give two examples of administrative controls (other than time, distance and local shielding) which can be used for the protection from external radiation hazards.

3. Define the following terms:

1. Controlled area.

2. Supervised area.

3. Unclassified area.

4. Once an area has been classified, what should be done to ensure that workers and visitors are aware of the classification in that particular area?

5. a)  What does an occupancy factor of 0.5 mean?

b)  Suggest average dose rates which could be used to classify areas where the occupancy factor is 0.5.

6. a)  Define the term investigation level.

b)  Why should investigation levels be used in radiation protection?

7. Give two examples of physical controls which can be used for the protection from external radiation hazards.

Now check your answers with the model answers your workbook.

 12.SHIELDING CALCULATIONS

The information in this section will enable you to make estimations of the thickness of shielding required for beta particles, x- and gamma rays.  This material is intended to be an introduction to shielding.  Further information for shielding in industrial and medical facilities (e.g dose rates from x-ray generators) is given in the industry specific modules in Part 4 of this distance learning course.  Neutron shielding calculations are complex and beyond the scope of this course.

 12.1Shielding of Beta Particles

The thickness of material required to effectively shield beta particles can be determined from the maximum range of the beta particles in the shielding material.  This can be estimated for water, polyethylene and perspex (with an assumed density of 1 g cm^-3) using the general rule that the distance travelled in centimetres equals the maximum beta energy in MeV divided by 2.  This relationship is shown in Equation 5:

d = E[5]

2

where   d = Distance travelled (in cm)

E = Maximum energy of the beta particles (in MeV)

!

This method of shielding beta particles is illustrated in Example 4.

Example 4

Question

Yttrium-90 has a maximum beta particle energy of 2.3 MeV.  What thickness of perspex shielding would you need to effectively shield this source?

Answer

E = 2.3 MeV

From Equation 5 above:

 d = 2.3 = 1.15 cm

2

As the distance travelled is equal to the shield thickness needed to absorb all the beta particles, then 1.15 cm of perspex would be needed as an effective shield for the beta radiation.

Although Example 4 provides a reliable estimate for the shielding of beta particles, it is important to remember that when beta particles interact with matter scattered x-rays, known as bremsstrahlung radiation, are produced.  Therefore, shielding of beta particles involves not only the shielding of the particles themselves, but also involves the minimization of bremsstrahlung radiation.  The fraction of beta particle energy converted to bremsstrahlung x-rays is given by Equation 6:

F = 3.3 x 10^-4 ZE[6]

where   F = Fraction of beta particle energy converted into x-rays

Z = Atomic number of the absorber/shielding material

E = Maximum energy of beta particles (in MeV)

Equation 6 indicates that the lower the atomic number of the shielding material, the lower the amount of bremsstrahlung x-rays produced.  Since many shielding materials are made out of compounds (e.g. perspex) rather than elements (e.g. lead), the value of Z is really an effective atomic number rather than a true atomic number.  Perspex, polythene and water are all assumed to have an effective atomic number of 6 or 7 for radiation protection purposes.  These materials are all suitable shielding for beta particles but some bremsstrahlung radiation will still be produced.  This is illustrated in Example 5.

Example 5

Question

What fraction of the beta particle energy from a yttrium-90 source (maximum beta energy of 2.3 MeV) is converted to x-rays in a perspex shield with an effective atomic number of 7?

Answer

E = 2.3 MeV

Z = 7

From Equation 6:

F = 3.3 x10^-4 x 7 x 2.3

F = 5 x 10^-3

F = 0.5%

For low activity sources, this quantity of bremsstrahlung x-rays produced will not be significant.  However, at activity levels in the order of megabecquerels (MBq), the bremsstrahlung radiation from a source such as yttrium-90 will require additional shielding.  This is usually in the form of a higher atomic number absorber (e.g. lead).  Hence, the most effective and commonly used shielding for energetic beta particles is perspex (to shield the betas) surrounded by lead (to shield the bremsstrahlung x-rays).

 12.2Shielding of X- and Gamma Rays

 12.2.1The attenuation of x- and gamma rays

When x- and gamma rays pass through material they are not totally absorbed by that material.  Instead they are attenuated (i.e. reduced in intensity).  X- and gamma rays are attenuated exponentially when they pass through material.  This attenuation can be represented mathematically by Equation 7:

R_x  = R_o e^-x[7]

where   R_x = Dose rate after passing through a shield of thickness x

  R_o = Dose rate without shielding

  x   = Shield thickness

    = A constant known as the linear absorption coefficient of the shielding material

The linear absorption coefficient depends on the type of shielding material used and also on the energy of the incident X- or gamma ray.  It is usually given in units of cm^-1 (i.e. the absorption per centimetre of shielding).

Although Equation 7 can be used to give a value for the amount of shielding material required, in practice simpler methods are used.  These methods use experimentally determined shielding quantities known as Half Value Layers and Tenth Value Layers.

 12.2.2Half value and tenth value layers

The half-value layer (HVL) (also known as the half-value thickness), for a particular shielding material, is the thickness of a shielding material required to reduce the intensity of radiation to half its original value.

Equation 7 is similar to the radioactive decay equation (A = A₀e^-T) and is used in a simpler form for shielding calculations using half value layers, in the same way that the radioactive decay equation is used in a simpler form using radiological half-lives.  The simpler form of Equation 7 is given by Equation 8:

R_x =  R₀  8

2ⁿ

where   n = the thickness of the shielding in terms of the number of HVLs

R_x = Dose rate after passing through a shield of thickness x

  R_o = Dose rate when unshielded

The actual thickness (x) of the shielding can be written in terms of Half Value Layers (HVL’s) as shown in Equation 9:

x =  nHVL9

The relationship between the linear absorption coefficient () and the half value layer (HVL) is represented by Equation 10:

HVL = 0.69310

 

Another useful shielding value is known as the tenth value layer (TVL).  This is defined as the thickness of a shielding material required to reduce the intensity of radiation to a tenth of its original value.  Equation 11 can be used with tenth value layers:

R_x =  R₀  11

10ⁿ

where   n = the thickness of the shielding in terms of the number of TVLs

R_x = Dose rate after passing through a shield of thickness x

R_o = Dose rate when unshielded

In this case, the actual thickness (x) of the shielding can be written in terms of Tenth Value Layers (TVL’s) as shown in Equation 9:

x =  nTVL12

The relationship between the linear absorption coefficient () and the tenth value layer (TVL) is represented by Equation 13.

TVL = 2.30313

 

Some typical values for HVL and TVL are shown in Table 5.  Table 5 gives the energies of the photons and the HVLs and TVLs  for concrete and lead.  Notice that the values of HVL and TVL increase with increasing X- or gamma ray energy.  This means that a thicker shield is required for a cobalt-60 source than for a caesium-137 source.  Note also that the HVL or TVL is larger for concrete than for lead.  This means that concrete shielding will need to be thicker than lead shielding.

Table 6

HVL and TVL Values for Several X- and Gamma Ray Sources

 Note: Values taken from NCRP Report No. 49 1976

The X-ray sources are described in terms of the maximum (or peak) voltage (in kilovolts) applied to the tube.  This is written as kV_p.  There will only be a very few x-rays at the energy corresponding to peak voltage and so the average x-ray energy will much less.

The values in Table 5 can be used to estimate the thickness of shielding required in order to reduce x- or gamma radiation dose rates to workers.  Example 6 illustrates how half value layers may be used in practice.

Example 6

Question

The dose rate near a cobalt-60 source is measured at 160 Sv h^-1.  What thickness of lead is needed to reduce this dose rate to 10 Sv h^-1?

Answer

This question can be answered using either HVLs or TVLs.  In this example half value layers will be used.  Using the HVL of lead for cobalt-60 from Table 5 we have the following data:

HVL  = 1.20 cm

R_x = 10 Sv h^-1

R₀ = 160 Sv h^-1

Using equation 8

 10 = 160

2ⁿ

 2ⁿ = 160

10

2ⁿ = 16

Since

16 = 2 x 2 x 2 x 2 = 2⁴

n = 4

Four HVLs of lead are required.

Hence, the thickness of lead needed is 4 x 1.2 = 4.8 cm.

Note that 2ⁿ  may not work out to be a number which is easy to process to obtain n.  In which case you will need to use logarithms in the same way as in Module 1.3 Ionizing Radiation and Radioactive Decay.

For the above example  

2ⁿ = 16

Using logarithms

nlog2 = log 16

(Note that common logs or natural logs can be used.)

 n = log 16

log 2

 n = 1.204

0.3010

 n = 4

Example 7 shows a similar calculation for tenth value layers.

Example 7

Question

The dose rate near a small caesium-137 source is measured at 40 Sv h^-1.  What thickness of concrete is needed to reduce this dose rate to 10 Sv h^-1?

Answer

This question can be answered using either HVLs or TVLs.  In this example tenth value layers will be used.  Using the TVL of concrete for caesium-137 from Table 5 we have the following data:

TVL  = 15.7 cm

R_x = 10 Sv h^-1

R₀ = 40 Sv h^-1

Using equation 8

 10 = 40

10ⁿ

 10ⁿ = 40

  10

10ⁿ = 4

Using logarithms

nlog10 = log 4

 n =  log 4  =  0.6021

log 10          1

n = 20.6021

Hence, the thickness of concrete needed is 0.6021 x 15.7 = 9.5 cm.

Remember that it is often more practical to use shielding to reduce exposure rather than using distance.  Example 8 shows a direct comparison between the two protection methods of distance and shielding.

Example 8

Question

1. For a caesium-137 source with a dose rate of 390 Sv h^-1 at 1 m, what would be the amount of lead shielding needed to reduce the dose rate to 25 Sv h^-1.

2. What distance would achieve this lower dose rate, if no shielding was available?

Answer

a)  For shielding:

From Table 5, the lead HVL of lead for caesium-137 is

HVL = 0.65 cm

R_x = 25 Sv h^-1

R₀ = 390 Sv h^-1

Using equation 8

 25 = 390

2ⁿ

 2ⁿ = 390

  25

2ⁿ = 15.6

Using logarithms

 nlog2 = log 15.6

 n = log 15.6

log 2

 n = 1.193

0.3010

n = 3.96

Alternatively, since shielding calculations are used to provide estimates of shielding requirements, 15.6 is approximately equal to 16.

Therefore,

2ⁿ = 16

n = 4

Approximately 4 HVLs are required.

Therefore the thickness of lead needed is 4 x 0.65 = 2.6 cm.

b)  For distance:

Using Equation 4 from Section 2.1.2 and substituting the values:

R₁ = 390 Sv h^-1

R₂ = 25 Sv h^-1

r₁  = 1 m

390 x (1² ) = 25 x r₂²

Then

r₂ = (15.6) = 3.95 m (approximately 4 m)

Hence, shielding at one metre with 2.6 cm of lead achieves the same reduced dose rate as moving four metres away from the source.  The latter option is not practical if work has to be performed on equipment near the source.  So, in practice the most efficient way of reducing exposure is to shield the source.

It is important to note that the calculations shown in this section deal with the theory of shielding x-and gamma rays.  In practice, others factors relating to the energy and geometry of the beam and type of shielding material will affect the amount of radiation transmitted.  In fact, these factors can result in an underestimate the amount of shielding required and hence higher dose rates than expected.  To overcome this problem, numerical multiplication factors known as build-up factors have been calculated for various energies, geometrys and shielding materials.  The theory and application of these factors is beyond the scope of this course but you should be aware that, although the above equations give an approximate estimation of the shielding required, practical considerations may mean that you could underestimate the necessary amount of shielding.  Hence, it is always a good idea to measure the dose rates after the shielding is in place to ensure that the required dose rate reduction has been achieved.

To ensure that dose rates are reduced in accordance with your estimates, you should always measure the dose rates after the shielding is in place.

If the measured dose rates are greater than expected then more shielding may be necessary.

PRACTICAL ACTIVITY 2

Measurement of Half Value Layers

In this practical assignment you will measure the effect of different thicknesses of various shielding materials on gamma dose rates.  You will then be able to estimate half value layers for the shielding materials you have used.  You should contact your supervisor to arrange a suitable time to do the practical.

SELF-CHECK 4

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

1. Phosphorus-32 has a maximum beta energy of 1.71 MeV.  What thickness of perspex shielding would you need to effectively shield this source?

2. a)  What are the two important factors which control the amount of bremsstrahlung produced in a shielding material?

b)  What fraction of the beta particle energy from a phosphorus-32 source (maximum beta energy of 1.71 MeV) is converted to x-rays in a perspex shield with an effective atomic number of 7?

3. Define the following terms:

a)  Half Value Layer (HVL).

b)  Tenth Value Layer (TVL)

4. a)  A cobalt-60 source gives a dose rate of 400 Sv h^-1 at 1 m.  At what distance from the source must a barrier be placed if the dose rate at the barrier must not exceed 25 Sv h^-1?

b)  Using the values for HVL in Table 5, calculate the thicknesses of the following materials which would give the same protection at the original distance of 1 m from the source:

i)  Lead.

ii)  Concrete.

Now check your answers with the model answers your workbook.

 13.CALCULATING GAMMA RAY DOSE RATES FOR DIFFERENT RADIONUCLIDES

 13.1Introduction

In many situations in radiation protection (e.g. when designing new plant or increasing the amount of radioactive materials kept in a store), it is useful to estimate what the dose rate would be at a particular distance from a  radioactive source.  For this reason, values known as specific gamma dose rate constants have been calculated and are available in radiation protection reference material.

 13.2Calculating Dose Rates Using The Specific Gamma Dose Rate Constant

The specific gamma dose rate constant (or specific gamma ray constant) is denoted by the Greek symbol  (the Greek upper case letter for ‘G’).  This value is constant for a particular gamma-emitting radionuclide and is defined as the unshielded dose rate (in Sv h^-1) at 1 metre per gigabecquerel (GBq) of activity.  Some examples of  for different radionuclides are shown in Table 6.

Table 7

Specific Gamma Dose Rate Constants for Some Common Radionuclides

 Note: Values for  are taken from Health Physics and Radiological

 Heath Handbook, 1992

Example 9 shows how the value of  may be used to calculate theoretical dose rates for a particular radionuclide.

Example 9

Question

What would the dose rate be at 1 m from a technetium-99m source with an activity of 100 MBq?

Answer

From Table 6:

 for Tc-99m is 33 Sv h^-1 per GBq at 1 metre.

That is, the dose rate at 1m from 1 GBq of Tc-99m is 33 Sv h^-1.

We have only 100 MBq of activity:

100 MBq = 0.1 GBq

So for 100 MBq of Tc-99m, we multiply  by 0.1:

33 x 0.1 = 3.3 Sv h^-1

Hence, the dose rate at 1 m from 100 MBq of Tc-99m would be 3.3 Sv h^-1.

In addition, the gamma ray dose constant for any radionuclide can be used in conjunction with Equation 4 in Section 2.1.2 to calculate dose rates at different distances.  This calculation is shown in Example 10.

Example 10

Question

What is the dose rate at 0.5 m from an iridium-192 source with an activity of 250 MBq?

Answer

From Table 6:

 for Ir-192 is 160 Sv h^-1.

That is, the dose rate at 1 m from 1 GBq of Ir-192 is 160 Sv h^-1.

We have only 250 MBq of activity:

250 MBq = 0.25 GBq

So for 250 MBq of Ir-192, we multiply  by 0.25:

160 x 0.25 = 40 Sv h^-1 at a distance of 1 m

Using Equation 4:

R₁ = 40 Sv h^-1

r₁ = 1 m

r₂ = 0.5 m

40 x 1² = R₂ x 0.5²

Rearranging this equation gives:

 R₂ =   40  = 160 Sv h^-1

0.25

Hence, the dose rate at 0.5 metre for 250 MBq of Ir-192 is 160 Sv h^-1.

SELF-CHECK 5

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

1. Define the term specific gamma dose rate constant.

2. What are the units of the specific gamma dose rate constant ()?

3. What is the dose rate at 1 m from an iodine-131 source of activity 260 MBq?

4. What is the dose rate at 5 m from a caesium-137 source of activity 10 GBq?

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

KEY POINTS

8. An external radiation hazard is described as a hazard from a radioactive source outside the body.

9. An internal radiation hazard exists when there is a potential for exposure from radioactive materials entering the body.

10. Beta particles, x-rays, gamma rays and neutrons are all external radiation hazards.

11. X-ray generators only present an external radiation hazard when they are turned on.

12. Radioactive materials represent a continuous radiation hazard.

13. External radiation hazards are controlled by applying three basic methods; those of time, distance and shielding.

14. The radiation dose received by a person working in an area and exposed to a certain dose rate depends on the time spent in the area.

15. If you halve the exposure time, the dose received is halved.  Conversely, if you double the exposure time, the dose received is also doubled.

16. The greater the distance from a source of penetrating radiation, the less the overall exposure.

17. When the distance from a radiation source is doubled, the dose rate is reduced to a quarter of its original value.

18. Shielding a source allows dose rates to be reduced whilst still allowing work to be performed.

19. The amount and type of shielding material depends on the type and energy of radiation, the activity of the source (or intensity of radiation from a machine) and the acceptable dose rate outside the shielding.

20. Alpha and low energy beta particles do not represent an external radiation hazard and therefore do not require shielding.

21. Perspex surrounded by lead (to stop bremsstrahlung) is recommended for shielding high energy beta particles.

22. Concrete, lead or iron are recommended for shielding x-and gamma rays.

23. Concrete, water , polythene or borated paraffin wax are recommended for shielding neutrons.

24. Administrative controls can be used in the workplace to minimize exposure to ionizing radiation.

25. Example of administrative controls include: classification of areas, warning signs, training, work procedures, local rules, source inventories, radiation safety auditing systems and investigation levels.

26. Work areas may be classified into controlled, supervised and unclassified areas according to the level of external radiation hazard.

• A controlled area is one where specific protection measures and safety provisions are or could be required to control normal exposures and to prevent or limit potential exposures.

• A supervised area is where occupational exposure conditions are kept under review but no specific protection measures are required.

• An unclassified area is where no protection methods are required and the occupational exposures do not need to be kept under review.

27. Occupancy time is the amount of time a worker spends in a particular classified area.

28. The occupancy factor is the fraction of work time spent in a particular classified area.  It is often written as a percentage.

29. The internationally accepted symbol for ionizing radiation is a black trefoil on a yellow background.

30. Investigation levels are set levels of a measured value.  For any measured value above this level, an investigation into the cause should take place.

31. Physical controls are essentially physical barriers which ensure that the chance of accidental exposure is as low as possible.

32. Example of physical controls are interlocks, fixed shielding, remote manipulators and preset timers.

33. The thickness of material required to effectively shield beta particles can be determined from the maximum range of the beta particles in the shielding material.

34. When x-and gamma rays pass through a matter they are attenuated exponentially.

35. The half-value layer (HVL) is the thickness of a material required to reduce the intensity of x- or gamma radiation to a half of its original value.

36. The tenth-value layer (TVL) is the thickness of a material required to reduce the intensity of x- or gamma radiation to a tenth of its original value.

37. Half value and tenth value layers can be used to estimate the thickness of material necessary to shield a particular source of ionizing radiation.

38. Due to certain practical considerations, you should always measure the dose rates after the estimated thickness of shielding is in place to ensure that the required dose rate reduction has been achieved.

39. The specific gamma dose rate constant () is constant for a particular gamma-emitting radionuclide and is defined as the dose rate in Sv h^-1 at 1 metre per gigabecquerel (GBq) of activity.

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 the assignment.

 GLOSSARY OF TERMS

DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 2.3

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 2

RADIATION PROTECTION

MODULE 2.3

PROTECTION FROM INTERNAL RADIATION HAZARDS

CONTENTS

 OVERVIEW

People who work with radioactive materials may take some of the radioactive material into their bodies.  An important part of radiation protection is to protect workers from the possible entry of radioactive materials into the body.

In this module you will learn how radioactive material can enter the body and how doses to body tissue can be assessed.  Most importantly, you will learn control methods which are used to reduce the chances of radioactive material entering the body.

 

TIME ALLOCATION

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

MATERIALS

You will need your workbook and writing materials to take notes and answer the questions in this module.  A scientific calculator would assist with the calculations.

NOW YOU ARE READY TO START WORK

 14.INTERNAL RADIATION HAZARDS

 14.1Introduction

You have learned about external radiation hazards in Module 2.2 Protection from External Radiation Hazards.  Remember that an external radiation hazard is due to potentially harmful ionizing radiation from a radioactive source outside the body.  An internal radiation hazard exists when there is a possibility of harm due to radioactive material entering the body.  Hence, an external radiation hazard could also be an internal radiation hazard if there is a possibility of the radioactive material entering the body.  The radiation protection methods for each type of hazard are different and this module covers the protection from internal radiation hazards only.  Figure 1 illustrates external and internal radiation hazards.

Figure 11

External and Internal Radiation Hazards

 14.2Unsealed Sources and Radioactive Contamination

Sources of radioactive materials can be divided into two types – sealed and unsealed sources.  A sealed source is defined as radioactive material that is either sealed in a capsule or is in a solid form whereby none of its radioactive material can be lost under normal use.  Conversely, an unsealed source is radioactive material that is not sealed or is in a form whereby some of the radioactive material could be lost under normal use (e.g. a powder, a liquid or a gas).  When radioactive material is sealed, it may present an external radiation hazard to the people working nearby.  However, when it is unsealed, it can presentan internal radiation hazard as there is a chance that this material may enter the body.  In addition, unsealed sources may also be an external radiation hazard, although usually to a lesser extent than sealed sources.

Radioactive material when released inadvertently or when present in any location where it is not wanted (including in or on the body) is defined as contamination.  If sources are incorrectly handled or involved in an accident, some of the radioactive material may be dispersed to the surrounding area and result in the radioactive contamination of equipment, benches, floors and/or personnel.  This contamination can either become securely fixed to a particular surface (fixed contamination), or it can be free to move (removable contamination).  The presence of removable contamination, either as surface contamination or airborne contamination, poses an internal radiation hazard as this radioactive material may enter the body and cause internal doses to body tissue.

It is important to remember that quite small quantities of contamination which represent an insignificant external radiation hazard may pose an appreciable internal hazard.  There are two main reasons for this:

1. Once radioactive material enters the body it will continue to irradiate the tissues and organs of the body until either the radioactive material has decayed or until the material is excreted.  You cannot walk away from the hazard or shield it, unlike an external radiation source.

2. Once radioactive material enters the body it is in contact with body tissue.  This is particularly hazardous for alpha emitting radionuclides such as americium-241 or plutonium-239.  These emitted alpha particles have a very small range in tissue (in the order of micrometres) and deposit all their energy in a very small volume.  Hence, the damage to the tissue is so great that the body cells are unlikely to be repaired.

As an example consider the potential doses from plutonium-239.  The external dose rate from 200 Bq of plutonium-239 is negligible.  However, if the same quantity of plutonium-239 were inhaled, this could lead to a committed effective dose of approximately 100 mSv (five times the annual effective dose limit recommended by ICRP Publication 60).

It is essential that the possibility of an intake of radioactive material is minimized as far as possible.

In general, only unsealed sources of radioactive materials are considered to pose an internal hazard but it should be noted that, if a sealed source is accidentally or deliberately breached, then it would also represent an internal hazard.

 14.3Practices Which Use Unsealed Sources

The use of unsealed sources in industry, science and medicine is common.  In industry and science, the use of radioactive tracer sources is well established and research experiments use unsealed sources as a common practice.  The most widespread use of unsealed sources is in nuclear medicine where the chemical composition of radioactive material and its carrier is carefully chosen so that they will go to a specific organ in the body.  The ability to detect the radiation from these sources makes them a useful diagnostic tool.  An example of this is the use of technetium-99m combined with various carriers for scans of the brain, liver, lungs, spleen and bone marrow.  In a similar way, iodine-131 can be used in therapy to kill malignant thyroid cells.  Sources which are used for medical purposes (i.e. diagnostic and therapeutic purposes) are known as radiopharmaceuticals.

SELF-CHECK 1

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

1. Define the following terms:

1. Sealed source.

2. Unsealed source.

3. Contamination.

2. What is the difference between removable and fixed contamination?

3. Why does the presence of removable contamination present an internal hazard?

4. Why may small quantities of contamination represent a significant internal hazard?

5. Which type of source (sealed or unsealed) is generally considered to pose an internal radiation hazard?

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

 14.4The Effect of Radiation Type on Internal Radiation Hazards

An important consideration in the protection from internal radiation hazards is the type of radiation emitted by the radioactive contamination.  The degree to which each type of radiation represents an internal radiation hazard is dependent on the range of each type of radiation in body tissue.

As already mentioned in Module 2.2 Protection from External Hazards, alpha particles pose no external radiation hazard due to their poor penetrating ability.  However, alpha radiation from radioactive material inside the body can damage body organs due to both the close proximity of the contamination to the organs or tissues and the short range of alpha particles in the body.  Hence internal alpha contamination deposits large amounts of energy over a very small volume of body tissue and this could result in damage to the organ or tissue surrounding the source.

Internal beta contamination may also pose a localized radiation hazard but to a lesser extent than alpha contamination.  In general, beta particles have a higher range than alpha particles in the body and hence their energy is transferred over a larger volume of tissue.  This in turn lessens the overall effect of the source on the surrounding organs and tissue.

Gamma emitting contamination represents less of an internal radiation hazard than either alpha or beta contamination.  This is due to the high penetrating ability and long range of this type of radiation in tissue.

Due to the nature of x-ray and neutron production, it would be rare to encounter internal x-ray or neutron contamination.

Table 1 summarizes the effect of radiation type on external and internal radiation hazards.

Table 8

The Effect Of Radiation Type On External And Internal Radiation Hazards

Type of Contamination	External Hazard	Internal Hazard
Alpha	None	Severe
Beta	Moderate	Moderate
Gamma	Severe	Slight

From Table 1 you can see that unsealed radioactive materials which emit alpha radiation are the most severe internal radiation hazard.  There must be very strict control over the use of unsealed alpha sources.  You will learn about methods of control later in this module.

 14.5How Radioactive Material Enters the Body

Contamination can exist in the form of a liquid, a gas or a powder (or dust) and this radioactive material can enter the body in several ways.  The four routes of entry are:

1. inhalation of airborne contamination;

2. ingestion through the mouth;

3. absorption through the skin; and

4. entry through an open wound.

Figure 2 represents the four routes by which radioactive material can enter the body.

Figure 12

Routes of Entry for Radioactive Material

When contamination is inhaled, a certain fraction will pass from the lungs into the bloodstream, another fraction will be exhaled and a third component will be swallowed when it is cleared from the lungs.  The relative size of these fractions depends on the chemical and physical form of the contamination and the physiology of the person involved.  Similarly, when contamination is ingested, the amount passing through the wall of the digestive tract and into body fluids depends on the nature of the contamination and the physiological conditions at the time of ingestion.

Some radioactive materials, such as tritiated water or radionuclides dissolved in organic solvents, are easily absorbed through the skin and can enter the body transport system.  In addition, contamination may also enter the bloodstream directly through an open wound.  Entry could be through a wound received whilst working with radioactive materials (e.g. a needlestick injury) or through a healing wound which may become contaminated during unsealed radioactive work.

SELF-CHECK 2

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

1. a)  Which type of radiation poses the most severe internal radiation hazard?

b)  Why is this so?

2. a) Which types of radiation pose the least severe internal radiation hazard?

b)  Why is this so?

3. State the four principle routes of entry of radioactive material into the body.

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

 15.MEASUREMENT OF INTERNAL RADIATION HAZARDS

When radioactive material enters the human body it may deliver radiation doses to various body organs and tissues.  In radiation protection, it is important to be able to assess the level of the hazard associated with an intake of radioactive material.  This is much harder to do accurately for the internal radiation hazard than for the external radiation hazard.  The effect of radiation on individual organs in the body is hard to determine because people have very different anatomical (size and shape), physiological (body processes such as respiration) and metabolic (changes such as rate of food digestion) characteristics.  In this section you will learn how assumptions are made in order to assess the level of internal radiation hazards.  

 15.1ICRP Reference Man and IAEA Reference Asian Man

There is a wide variation in the anatomical, physiological and metabolic characteristics of human beings.  For the purposes of radiation protection, ICRP Publication 23 defined Reference Man to describe the characteristics of a human being.  Reference Man was chosen to represent an average person over a wide range of human characteristics and is intended for use as a model for calculations and comparisons of the internal effects of ionizing radiation.  The data is currently being updated using studies of European and North American populations.  Appendix A gives some characteristics of Reference Man.

However, Reference Man is not a good representation of an Asian population and therefore equivalent data for Reference Asian Man have been produced by the International Atomic Energy Agency (IAEA).

 15.2Biological and Effective Half-Lives

Whenever an intake of radioactive material occurs, it is important to have some idea of how long the material will remain in the body in order to assess the hazard.  As already discussed in Module 1.3 Ionizing Radiation and Radioactive Decay, radioactive materials decay exponentially with time.  However, other bodily processes (e.g. urination and defecation) also act to remove these materials from our body and the rate at which the body excretes radioactive material is also approximately exponential.  In order to calculate the effective rate at which radioactive materials decay or are removed from the body, we need to combine both the rate of radioactive decay and also the rate of excretion.  In radiation protection the term effective decay constant (_eff) is used to describe the rate of decay and removal of radioactive materials from the body.  A mathematical definition of this term is given in Equation 1.

_eff= _r + _b[1]

where  _eff  = effective decay constant

  _r  =  radioactive decay constant

  _b  =  biological decay constant

Equation 1 can be related to effective, biological and radioactive half-lives (and hence time) by remembering that the radioactive decay constant can be written as:

 _r = ln 2

  T_r

where   _r = radioactive decay constant

  T_r = half-life of the radioactive material

  ln 2 = log_e 2 = 0.693

As both the effective decay constant and biological decay constant can be written in a similar way, Equation 1 can be rewritten as Equation 2:

1   =  1  +  1[2]

T_eff        T_r        T_b

where   T_eff =  effective half-life of a radionuclide in the body (i.e. the effective time taken for the activity of a radionuclide in the body to be reduced to half its original value by the combination of radioactive decay and biological excretion).

T_r  =   radioactive half-life of a radionuclide in the body (i.e. the time taken for the activity of a radionuclide in the body to be reduced to half its original value by radioactive decay).

T_b  =  biological half-life of a radionuclide in the body (i.e. the time taken for the activity of a radionuclide in the body to be reduced to half its original value by biological excretion).

Hence, if the values for T_r and T_b are known, then it is possible to calculate T_eff from Equation 2.

The radioactive half-life is a physical constant which depends only on the radionuclide.  The biological half-life (and hence the effective half-life) is dependent on the chemical and physical nature of the radioactive contamination and on the anatomical, physiological and metabolic nature of the person. If the radioactive material is chemically similar to other elements in the body, it can change places with them.  For example, strontium has similar chemical properties to calcium, so strontium-90 can replace calcium in bones.  The physical form of the contaminant is important as it determines the characteristics of the material, such as solubility, which affect how the material is absorbed into the body.

Example 1 illustrates how Equation 2 is used to calculate the effective half-life of radioactive contamination in the body.

EXAMPLE 1

Question

Calculate the effective half-life for cobalt-60 in body tissue if the radioactive half-life is 1.9 x 10³ days and the biological half-life is 10 days.

Answer

Using Equation 2:

1   =       1        +   1

T_eff        1.9 x 10³10

Hence:

T_eff = 9.95 days (approximately 10 days)

!

It can be seen from Example 1 that if the biological and radioactive half-lives differ greatly, then it is the smaller number that determines the effective value.

Table 2 gives some examples of radioactive, biological and effective half-lives for different radionuclides.

Table 9

Radioactive, Biological and Effective Half-Lives For Some Radionuclides

Radionuclide	Critical Organ	Tr (days)	Tb (days)	Teff (days)
60Co	Total body	1.9 x 103	10	10
90Sr	Total body	1.1 x 104	1.1 x 104	5.5 x 103
137Cs	Total body	1.1 x 104	70	70

Values taken from Medical Effects of Ionizing Radiation, Mettler and Moseley, 1985

SELF-CHECK 3

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

1. Define the following terms:

1. Effective half-life.

2. Biological half-life.

3. Radioactive half-life.

2. Which factors determine the biological half-life (and hence the effective half-life) of a particular radionuclide?

3. Calculate the effective half-life for tritiated water in body tissue if the radioactive half-life is 4.5 x 10³ days and the biological half-life is 12 days.

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

 15.3Dose Coefficients

The level of the internal radiation hazard depends on the size of the dose that occurs after radioactive contamination has entered the body.  Dose coefficients for Reference Man have been determined using Reference Man data and models of the human body which describe how the body works.  Dose coefficients (h) are measured in Sv Bq^-1 and give the dose to the whole body (committed effective dose) resulting from an intake of radioactive material.

The dose coefficient (h) is defined as the committed effective dose per becquerel of intake.

The committed effective dose per becquerel of intake for a particular nuclide is dependent on the method of intake.  The values for h for ingestion and inhalation for different radionuclides have been determined for occupationally exposed workers and are listed in ICRP Publication 68 and in Table II-III of the IAEA Basic Safety Standards.

Values of h for members of the public which take into account different age groups, have also been published in the IAEA Basic Safety Standards (Tables II-VI and II-VII) and in ICRP Publications 71 and 72.

Note that, as yet, no dose coefficients exist for Reference Asian Man.  Hence dose coefficients for Reference Man are used both in the IAEA Basic Safety Standards and throughout this module.

 15.4Annual Limit On Intake

The internal radiation hazard can be assessed by comparing the amount of radioactive material that could be taken into the body with the amount which would result in a dose equal to the relevant annual limit (currently, this dose limit for occupationally exposed workers is 20 mSv).  This gives us a quantity known as the Annual Limit on Intake (ALI).

One ALI is defined as the number of becquerels of a radionuclide which, if taken into the body by Reference Man, would result in a committed effective dose equal to the relevant ICRP recommended dose limit.

Mathematically, the ALI (in becquerels) for a particular radionuclide is calculated using Equation 4:

ALI = L[4]

h

where   L = the relevant annual effective dose limit.

h = the committed effective dose per becquerel of intake (in units of Sv Bq^-1).

An example of how to calculate ALIs for occupationally exposed personnel is given in Example 2.

EXAMPLE 2

Question

Calculate the ALI for inhalation for an occupationally exposed worker who is using unsealed phosphorus-32 where the dose coefficient h (inhalation) is 2.9 x 10^-9 Sv Bq^-1.

Answer

The relevant dose limit for an occupationally exposed worker is 20 mSv, i.e. 0.02 Sv.

Using Equation 4:

 ALI (inhalation) =      0.02     = 6.9 x 10⁶ Bq

 2.9 x 10^-9

Hence, an intake by inhalation of 6.9 x 10⁶ Bq of phosphorus-32 is equivalent to one ALI and, hence, is also equivalent to an annual committed effective dose of 20 mSv.

Some examples of ALIs for occupationally exposed workers are given in Table 3.

Table 10

Annual Limits on Intake For Radiation Workers For Selected Radionuclides

Radionuclide	ALI for Inhalation* (Bq)	ALI for Ingestion (Bq)
32P	6.9 x 106	8.3 x 106
60Co	1.2 x 106	5.9 x 106
90Sr	2.6 x 105	7.1 x 105
99mTc	6.9 x 108	9.1 x 108
131I	1.8 x 106	9.1 x 105
137Cs	3.0 x 106	1.5 x 106

 *  These values are based on values recommended in the IAEA Basic Safety Standards with the slowest possible absorption from the lungs.

The dose received by a person following an intake of radioactive material may be quoted in terms of a fraction of an ALI or in units of millisieverts.  Example 3 illustrates this concept.

EXAMPLE 3

Question

A radiation worker accidentally ingests 1.48 x 10⁶ Bq of cobalt-60.  Estimate the dose received by the worker (using the Reference Man model);

a)  in terms of ALI.

b)  in millisieverts.

Answer

Using Table 3:

5.9 x 10⁶ Bq = 1 ALI for ingestion

So:

 Number of ALIs ingested = 1.48 x 10⁶ = 0.25 ALI

  5.9 x 10⁶

As 1 ALI is equivalent to a committed effective dose of 20 mSv, then 0.25 ALI is equivalent to a committed effective dose of 5 mSv.

Therefore, the dose received can be quoted in terms of:

a)  0.25 ALI or

b)  5 mSv.

SELF-CHECK 4

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

1. In which documents can the dose coefficients be found for occupationally exposed workers?

1. Define the term annual limit on intake (ALI).

1. Calculate the ALI for an occupationally exposed worker for ingestion of americium-241 where the dose coefficient for ingestion is 2 x 10^-7 Sv Bq^-1.

1. A radiation worker accidentally ingests 2.25 x 10⁶ Bq of caesium-137.  Estimate the committed effective dose received by the worker (using the Reference Man model);

a)  in terms of ALI.

b)  in millisieverts.

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

 16.CONTAMINATION LIMITS IN THE WORKPLACE

The internal radiation hazard in the workplace is assessed by measuring levels of surface and airborne contamination and comparing these measured values to potential doses.  This comparison is done for surface contamination by using a quantity known as the Derived Limit (DL) for surface contamination and for airborne contamination using a quantity called the Derived Air Concentration (DAC).

 16.1Derived Limits for Surface Contamination

Derived Limits (DLs) can be used to assess and control the hazard from surface contamination in the workplace.  They are quoted in terms of becquerels per square centimetre (Bq cm^-2) for ease of comparison with measured levels, and are often used to classify work areas (as discussed in Section 4.2.1 of this module).

The method of determining Derived Limits (DLs) is quite complex but essentially they are calculated by considering the highest potential dose that could result from the skin contamination, inhalation or ingestion of a particular radionuclide, and relating this value to the relevant ICRP dose limit.  Although Derived Limits are not internationally defined, they are usually based on one times the total dose limit.  For example, in Australia, exposure to one Derived Limit (DL) for 2000 hours (8 hours a day, 5 days a week, 50 weeks per year) is deemed to result in a committed effective dose of 20 mSv (i.e. one times the ICRP dose limit for an occupationally exposed worker).  However, in other countries, exposure to one Derived Limit (DL) for 2000 hours could be taken to mean a fraction or multiple of the ICRP dose limit.  Hence, when referring to Derived Limits is important that you know the definition of a Derived Limit for that particular country or state.

When referring to Derived Limits (DLs), it is important that you know the correct definition of that limit for the particular country or state you are in.

The potential dose that a particular radionuclide can deliver to a human when taken into the body depends on the amount of harm that the radionuclide can cause.  This is known as the radiotoxicity of a particular radionuclide and it is dependent on:

• the type of radiation emitted by the radionuclide;

• its effective half life; and

• the radiosensitivity of the body organs it is likely to target.

To simplify the use of Derived Limits (DLs) in radiation protection, radionuclides with similar radiotoxicity are grouped together.  Appendix B gives the radionuclide groupings and corresponding Derived Limits for surfaces in classified areas as recommended by the National Health and Medical Research Council of Australia.  An example of a radionuclide in each group is given in Table 4.

Table 11

Radiotoxicity and Derived Limits for Surface Contamination

Group	Radiotoxicity	Example	Derived Limit* (Bq cm-2)
1	Very High	Am-241	0.1
2	High	U-238	1
3	Higher Medium	Sr-90	10
4	Lower Medium	I-131	100
5	Low	Tc-99m	1000

*  Based on an annual dose limit of 20mSv as recommended by ICRP 60.

Note that Group 1 and Group 2 radionuclides are alpha emitters and are classed as highly radiotoxic due to the severe effects of an intake.  Hence, the DL values for this group are much lower than any other group.

Example 4 shows how it is possible to compare and assess the hazard from surface contamination by using Derived Limits (DLs) as a guide.

EXAMPLE 4

Question

On a routine survey of an iodine-131 laboratory, you find a removable surface contamination level of approximately 50 Bq cm^-2.  What is this level in terms of DLs and what is the annual committed effective dose which could result from the exposure to this surface contamination level?

Answer

From Table 4, the DL for iodine-131 (I-131) is quoted as 100 Bq cm^-2 by the National Health and Medical Research Council of Australia.

Hence, the measured level is equivalent to (50  100) 0.5 DL’s.

According to the Australian definition of a Derived Limit, exposure to 1 DL for 2000 hours (8 hours per day, five days a week, 50 weeks a year) would result in a committed effective dose of 20 mSv.

Hence, we can say that exposure to 0.5 DL’s for 2000 hours would result in a committed effective dose of (0.5 x 20 mSv) 10 mSv.

Obviously, this surface contamination will decay away with time but, in accordance with the ALARA principle, levels of this magnitude should be reduced as far as possible.  However, this example shows how, by quoting surface contamination levels in terms of Derived Limits, we can assess the level of internal hazard.

Note that for comparison purposes, surface contamination measurements are usually averaged over certain areas.  For example, skin contamination is averaged over 100 cm² (except for the hand where the whole hand is averaged).  For walls, floors and ceilings, the contamination is averaged over 1000 cm².  The usual area for other surfaces is 300 cm².

 16.2Derived Air Concentration

Derived Air Concentrations (DACs) are used in the workplace to compare measured airborne contamination levels with levels which could result in an intake of one ALI.

One DAC is the concentration of a radionuclide in air (in Bq m^-3) that would result in a worker inhaling one ALI in a year.

The Reference Man model is used which gives a breathing rate of 2 400m³ per year (i.e. a nominal breathing rate of 1.2m³ per hour for an 8 hour day, 5 days per week and 50 weeks per year).

The mathematical definition for DAC is given in Equation 5:

DAC =  ALI (inhalation) [5]

2 400

An example of how to calculate DACs for work areas is shown in Example 5.

EXAMPLE 5

Question

Using the annual limits on intake given in Table 3, calculate the Derived Air Concentration (DAC) for phosphorus-32.

Answer

From Table 3:

ALI (inhalation) = 6.9 x 10⁶ Bq

Using Equation 5:

 DAC = 6.9 x 10⁶  = 2.875 x 10³ Bq m^-3

2400  

Hence the DAC for phosphorus-32 is 2.875 x 10³ Bq m^-3

Example 6 shows how the internal hazard can be assessed by comparing measured airborne activity with Derived Air Concentrations (DACs) and hence recommended dose limits.

EXAMPLE 6

Question

The airborne contamination in a laboratory which deals solely with iodine-131 is measured as 75 Bq m^-3.  What is this airborne contamination level in terms of Derived Air Concentrations (DACs) and what is the annual committed effective dose associated with this airborne contamination level?

Answer

From Table 3:

ALI (inhalation) = 1.8 x 10⁶ Bq

Using Equation 5:

 DAC = 1.8 x 10⁶  = 7.5 x 10² Bq m^-3= 750 Bq m^-3

2400

The airborne contamination in the laboratory is measured as 75 Bq m^-3.

Hence, the airborne contamination in terms of Derived Air Concentrations is:

 75  =  0.1 DAC

750

Remembering that 1 DAC is the concentration of a radionuclide which would result in inhaling 1 ALI in a year, then 0.1 DAC would result in inhaling 0.1 ALIs in a year.  Also remembering that an intake of 1 ALI would result in an intake of 20 mSv.  Then an intake of 0.1 DACs would result in a committed effective dose of (0.1 x 20) 2mSv.

Hence, by using DAC levels as a guide, it is possible to compare and assess the internal radiation hazard from airborne contamination for particular work areas.  In addition, DAC levels may be used to classify work areas in terms of their internal radiation hazard.  This concept will be further discussed in Section 4.2.1 of this module.

SELF-CHECK 5

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

1. What are the three main uses of Derived Limits in the workplace?

2. a)  Which type of radionuclides are the most radiotoxic?

2. Why is this so?

3. On a routine survey of a sulphur-35 (S-35) production facility, you find a removable surface contamination level of approximately 250 Bq cm^-2.  a)  What is this surface contamination level in terms of Derived Limits?

b)  What is the annual committed effective dose which could result from the exposure to this surface contamination level?

4. Define the term Derived Air Concentration (DAC).

5. Using the Annual Limits on Intake given in Table 3, calculate the Derived Air Concentration for:  

a)  cobalt-60.

b)  technetium-99m.

6. The airborne contamination in a laboratory which deals solely with phosphorus-32 is measured as 5.8 x 10² Bq m^-3.

1. What is this airborne contamination level in terms of Derived Air Concentrations?

2. What is the annual committed effective dose associated with this airborne contamination level?

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

 17.CONTROLLING INTERNAL RADIATION HAZARDS

There are three basic techniques to consider when controlling the internal radiation hazard in a particular area:

1. Minimization of the amount of activity being handled;

2. Administrative controls; and

3. Physical controls to contain the activity.

 17.1Minimization of Radioactive Materials

In all processes involving the handling of radionuclides, the amount of activity should be reduced as far as possible.  This is in accordance with the ALARA principle.  In addition to reducing the hazards due to contamination, minimization of the activity also reduces the external dose rates to workers.

As an example, consider the situation in which radioactive material is being used in a tracer experiment.  The minimum amount of activity used will depend on the minimum amount that can be detected by the measuring equipment and the amount of activity that may be lost before reaching the measuring equipment.  Example 7 shows how to work out the minimum activity to use.

EXAMPLE 7

Question

An experiment is being conducted using phosphorus-32 (P-32) to see what fraction of phosphorus in the soil is taken up into the leaves of a certain plant.  Mathematical modelling predicts that 10% of P-32 will be taken up.  For statistical accuracy 1 kBq is the minimum activity to be detected by the measuring equipment.  Previous experiments have indicated an error factor of 2.  What minimum activity of phosphorus-32 would you recommend?

Answer

Minimum activity to be detected = 1 kBq

Predicted take up factor = 10%

Therefore:

Initial activity = 10 x 1 kBq = 10 kBq

To allow for an error of 2:

Minimum activity recommended = 2 x 10 kBq = 20 kBq

 17.2Administrative Controls

An administrative control is a method of administration which prevents or minimizes exposure to radiation hazards.  Administrative controls for internal radiation hazards are very similar to those for external radiation hazards, although the emphasis must be on preventing radioactive material from entering the body.  The following list gives some useful examples of administrative controls:

8. Classification of areas.

9. Display of clear signs for each classified area.

10. Radiological protection training for workers and managers.

11. Work procedures which incorporate the use of physical controls.

12. Local rules, (e.g. restriction of access to certain areas) and conditions of work, (e.g. requirement to wear protective clothing and respiratory protection).

13. Maintenance of source inventories for each area.

14. Radiation safety auditing system which includes safety assessment of workplace plant and equipment.

15. Use of investigation levels for personal dose monitoring and workplace monitoring results.

The classification of areas and local rules which have particular relevance to the internal radiation hazard will be discussed in the next two sections.  Other methods of administrative control will be discussed in the relevant modules of Parts 3 and 4 of this distance learning material.

 17.2.1Classification of work areas

Work areas can be classified according to the level of the internal radiation hazard.  A classification system in which areas may be classified as controlled or supervised is recommended in the IAEA Basic Safety Standards.  The definitions given here are the same as those given in Module 2.2 Protection from External Radiation Hazards.

A controlled area is one where specific protection measures and safety provisions are or could be required to control normal exposures and to prevent or limit potential exposures.

A supervised area is where occupational exposure conditions are kept under review but no specific protection measures are required.

An unclassified area is where no protection methods are required and the occupational exposures do not need to be kept under review.

Generally, the classification of areas varies in accordance with the particular needs of the workplace and is usually based on operational experience and judgement.  If the major radiological hazard is due to external radiation then the classification system suggested in Module 2.2 Protection from External Radiation Hazards may provide sufficient control of the internal radiation hazard as well.  If however the external radiation hazard is small compared with the internal radiation hazard, it is useful to classify areas according to the potential for airborne and surface contamination.  The area classifications are determined by considering what radionuclides are being used, how they are being used and the maximum activity used at any one time.  Table 5 illustrates one example of how the DL and DAC values can be used to classify areas which have an occupancy factor of 1 (i.e. a 100% occupancy of the classified area).  Of course, in practice the levels should be set locally depending on the conditions in the working areas and on the occupancy factor.

Table 12

Area Classification In Terms of Surface and Airborne Contamination Levels

 17.2.2Local Rules

Local rules are used in the workplace to control contamination levels and reduce the chance of radioactive material being taken into the body.  The following list gives some examples of subjects which should be covered in local working procedures:

• Prohibition of eating, drinking, smoking or applying cosmetics in classified areas.

• Maintenance of good housekeeping procedures.

• Use of protective clothing and respiratory protection.

• Regular area and personnel contamination monitoring.

• Controlling access to classified areas.

• Procedures for entering and leaving classified areas.

• Maintenance of source inventories.

 17.3Physical Controls

Although administrative controls provide a basis for a radiation protection program, these controls rely on familiarity with written procedures.  If a new worker is not aware of these procedures, or these procedures are not followed correctly for any other reason, then the potential exists for accidental exposure to internal radiation hazards.  In order to ensure that the chance of accidental exposure is as low as possible, physical controls are employed in the workplace to prevent the spread of radioactive materials.  The following list gives some useful examples of physical controls:

• Splash trays.

• Barriers.

• Fume cupboards.

• Glove boxes.

• Shielded cells.

• Interlocks.

 17.3.1Containment Systems

Containment systems consist of physical structures, (which may be as simple as a splash tray or as complex as a system of sealed containers with associated ventilation), which contain the radioactive materials and prevent the spread of contamination.  The choice of equipment for the effective containment of radioactivity depends on the quantity of activity and the physical form of the radioactive material.  As physical form may change during various processes (e.g. by boiling a liquid to produce the vapour), it is important to be aware of all operations being performed in a particular area when designing suitable containment systems.

Most containment systems for unsealed sources have at least two levels of containment.  Figure 3 shows four levels of containment suitable for handling radioactive liquids in a laboratory.

Figure 13

Four Levels of Containment For Handling Radioactive Liquids

Greater levels of containment are needed to control airborne contamination hazards such as vapour and dust.  Gloveboxes (i.e. boxes with fitted gloves) and shielded cells with remote manipulators are types of containment which allow handling of radioactive materials without any contamination coming into contact with the body.  Gloveboxes such as that shown in Figure 4 should only be used for handling alpha or beta emitters as they are usually made of perspex and have little or no shielding to protect against the external hazard.

Figure 14

Schematic Drawing of a Glove Box

For high dose rate gamma emitters with airborne or surface contamination hazards, a shielded cell such as that shown in Figure 5 can be used.  This is a cell with a thick lead glass shielding window and ‘master-slave’ manipulators for indirect handling of unsealed radionuclides.

Figure 15

Work Being Conducted in a Shielded Cell

The ventilation systems for glove boxes and cells usually involve inlet and exhaust filters.  The atmosphere inside is kept at a lower pressure than that outside so that should a leak develop, air will flow into the enclosure and prevent contamination from escaping.

A lower level of containment is provided by a fume cupboard (as shown in Figure 6) where the material is handled through an opening in the front.  Air is drawn from the laboratory through the front opening and into the fume cupboard.  This is then expelled through exhaust outlets and may be filtered before entering the environment.  The controls for services such as electricity and ventilation are positioned outside the fume cupboard to reduce the number of times the airflow through the front opening is disturbed.  It is important that fume cupboards are used correctly to ensure that contamination is being controlled.  This means that equipment in the fume cupboard should be minimized (no unnecessary storage), the area around the front of the fume cupboard should be kept free of obstruction, air flow at the face of the fume cupboard should be monitored and the opening at the front of the fume cupboard should be kept to a minimum.

Figure 16

A Fume Cupboard

 17.3.2Control of Radioactive Waste

Good containment of radioactive waste materials is an important aspect of contamination control and is often overlooked.  You will learn about safe methods of storing and handling radioactive waste in Module 3.4, Safe Handling of Radioactive Waste.

 17.3.3Laboratory Design

Design features of a laboratory in which unsealed radioactive materials are used may also be considered as physical controls.  It is important that surfaces which may become contaminated (bench tops, floors, walls etc.) are easy to clean, and that there are no absorbent surfaces or cracks which could trap contamination.  If you work in a radioisotope laboratory, or are involved in designing suitable laboratories, you will learn more about this in Module 4.3 Safety in Radioisotope Laboratories.

SELF-CHECK 6

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

1. What are the three basic principles for the control of internal radiation hazards?

2. Give three examples of administrative controls for the protection of individuals against the internal contamination hazard.

3. Give three examples of local rules which help control the internal radiation hazard.  

4. Describe what is meant by a containment system and give four examples of containment in a laboratory.

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

KEY POINTS

16. An internal radiation hazard exists when there is a possibility of harm due to ionizing radiation from a radioactive source entering the body.

17. Unsealed radioactive material in an undesirable location is known as contamination.

18. A sealed source is defined as radioactive material that is permanently sealed in a capsule or is in a solid form which cannot conceivably lose any of its material.

19. An unsealed source is defined as radioactive material that is not sealed or is in a form whereby it could lose some of its material (e.g. a liquid or gas).

20. Removable contamination is surface or airborne contamination which can move freely.  Fixed contamination is attached to a particular surface.

21. Quite small quantities of contamination which represent an insignificant external radiation hazard may pose an appreciable internal hazard.

22. In general, only unsealed sources of radioactive materials are considered to pose an internal hazard but, if a sealed source is accidentally or deliberately breached, then it would also represent an internal hazard.

23. Alpha radiation represents the most severe internal radiation hazard due its relatively high energy and short range in organs and tissue.

24. Unsealed radioactive materials can enter the body through inhalation, ingestion, absorption through the skin and through a wound.

25. The length of time a radionuclide stays in the body depends on the chemical and physical form of the contamination and the physiology of the person involved.

26. The ICRP recommends Reference Man as a model for internal radiation hazard calculations.

27. The term effective decay constant (_eff) is used to describe the rate of decay and removal of radioactive materials from the body.

28. The effective half-life (T_eff) of a radioactive material in the body is the time taken for the radioactive material in the body to be reduced to half its original value by the processes of radioactive decay and biological excretion.

29. The radioactive half-life (T_r) of a material is the time taken for the radioactive material to be reduced to half its original value by radioactive decay.

30. The biological half-life (T_b) of a radioactive material in the body is the time taken for the radioactive material in the body to be reduced to half its original value by biological excretion.

31. Dose conversion coefficients are used to calculate the amount of activity taken into the body to a committed effective dose.

32. The annual limit on intake (ALI) is defined as the intake of a particular radionuclide by inhalation, ingestion or absorption through the skin or wound such that would result in an annual committed effective dose equal to the relevant dose limit.

33. Derived Limits (DL) are limits of surface contamination such that DACs, ALIs and skin dose limits are not exceeded.

34. The Derived Air Concentration (DAC) is defined as the concentration of a particular radionuclide that would result in an occupationally exposed person inhaling one ALI in a year.

35. Areas may be classified in terms of the internal radiation hazard (DACs and DLs).

36. The internal hazard can be controlled by minimization of radioactive materials, administrative controls and physical controls.

37. Local rules are necessary in areas where unsealed radioactive material is used.

38. Containment systems are physical structures which are designed to prevent the spread of radioactive contamination.

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 the assignment.

 APPENDIX A

SOME CHARACTERISTICS OF REFERENCE MAN

Some Organs of Reference Man (Adult Male)

Organ	Mass (kg)	% of Total Body
Total Body	70	100
Skeleton	10	14
Muscle	28	40
Fat	13.3	19
Blood	5.5	7.9
Gastrointestinal Tract	2.2	3.1
Thyroid Gland	0.02	0.029

Water Balance

Intake Method	Litres/day	Excretion	Litres/day
Foods	0.70	Urine	1.40
Fluid	1.95	Sweat	0.65
Oxidation of Food	0.35	Insensible*	0.85
Total	3.00	Faeces	0.10

 *In exhaled air, and as a result of diffusion through the skin and a very low sweat rate

Air Inhalation Volumes

Air inhaled during 8 hour working day	9.6 m3
Air inhaled during 16 hours not at work	13.2 m3

 APPENDIX B

SURFACE CONTAMINATION LIMITS

Taken from Recommended limits on radioactive Contamination on Surfaces in Laboratories (1995), Radiation Health Series No. 38, Australian National Health and Medical Research Council.

 GLOSSARY OF TERMS

DISTANCE LEARNING MATERIALS

RADIATION PROTECTION

Module 2.4

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 2

RADIATION PROTECTION

MODULE 2.4

USE OF RADIATION MONITORING INSTRUMENTS

CONTENTS

 OVERVIEW

An important part of a radiation protection program is the monitoring of areas where radiation hazards may exist.  Levels of the radiation hazard can be assessed by performing dose rate surveys (for the external hazard) and contamination surveys (for the internal hazard).  In this module you will learn how to select appropriate instruments for measuring radiation hazards and how to use them effectively.  You will also learn how to perform radiological surveys safely for dose rates and contamination, and how to report the results in a meaningful way.

Before proceeding with this module, you may wish to revise Module 1.5 Methods of Radiation Detection.

 

TIME ALLOCATION

It is estimated that you will need to spend ten to twelve hours to complete this module.  This time estimate is only for guidance only.  Do not worry if you need to spend more time to understand this module.

MATERIALS

You will need writing materials to take notes and answer questions.  Access to a scientific calculator is also desirable.

NOW YOU ARE READY TO START WORK

1. 
   RADIATION MONITORING INSTRUMENTS

1.  Types of Detectors

Radiation detectors are the part of the instrument which interacts with incoming radiation and converts the energy into something (usually light or electric charge) which can be measured.  By studying Module 1.5 Methods of Radiation Detection, you have gained basic knowledge about how the following types of detector work:

• Gas-filled detectors (ionization chambers, proportional counters, GM counters);

• Solid state conductivity detectors; and

• Scintillation detectors.

2.  Types of Instruments

Radiation measuring instruments must be able to detect and quantify external exposure from radiation sources outside the body and detect and quantify contamination which may enter the body and cause internal exposure.  Hence, there are two types of measuring instrument which are used in the workplace:

• Dose rate meters (for the external hazard); and

• Contamination monitors (for the internal hazard).

As you may remember from Module 1.5 Methods of Radiation Detection, other instruments (known as spectrometers) can be used for spectroscopy purposes (i.e. to separate out energies and hence identify radionuclides).  As the operation of these instruments is more complex, they are not covered in this module.

1.  Dose rate meters

Dose rate meters contain a detector capable of detecting penetrating radiation and some beta particles.  Two typical dose rate meters are shown in Figure 1.

Geiger-Müller Tube Dose Rate Meter

Ion Chamber Dose Rate Meter

Figure 17

Two Typical Dose Rate Meters

Most dose rate meters can measure both gamma and x-ray dose rates, and those with thin end windows can also measure beta dose rates.  As you should remember from Module 2.2 Protection from External Radiation Hazards, alpha particles cannot penetrate the skin and therefore do not present an external radiation hazard.  Hence, dose rate monitoring for these particles is unnecessary.  Neutrons are more difficult to monitor because they do not cause ionization directly.  It is therefore usual to measure neutron dose rates with specially designed neutron monitors (as discussed in Module 1.5 Methods of Radiation Detection).  As monitoring for neutron radiation is less common for routine surveys, the use of these specialised neutron monitoring instruments is not discussed in this module.

Dose rate meters usually give a direct reading of dose rate in units of microsieverts per hour (Sv h¹) (see Figure 2).

Figure 18

Monitoring for Dose Rate

The most popular unit on older dose rate meters is mrem h^-1 and the easiest way to convert from mrem h^-1 to Sv h^-1 is to remember that:

1 mrem h^-1 = 10 Sv h^-1

For example, if a dose rate monitor showed a dose rate of 0.75 mrem h^-1,this is equivalent to 7.5 Sv h^-1 (i.e. 0.75 x 10 Sv h^-1).

Some specialised dose rate meters used for dosimetry or environmental monitoring may give readings in units of micrograys per hour (Gy h^-1).  Although for these specialist uses it may be necessary to use factors which depend on the radiation energy and the material being exposed, for most safety related measurements of dose rate from gamma or x-ray sources, it is usual to assume grays (Gy) and sieverts (Sv) are the same.

For most safety related measurements of dose rate from gamma or x-ray sources:

1 mGy = 1 mSv

A brief overview of these units is given in Reference Material R1.  They will also be discussed more fully in Module 2.5 Personal Dosimetry.

Note that the detectors in dose rate meters are actually measuring the effect of ionizing radiation on the detecting material.  However, in radiation protection, we are more concerned with the effect of the radiation on a person.  In fact, if you remember from Module 2.1 Principles of Radiation Protection, dose limits are given in terms of effective dose per year (a quantity, explained in Module 2.5, used to describe the dose received by a person).  So, in order to estimate the effective dose due to a particular source, we would need to know the effect in all the tissues and organs of the body.  Of course, this is very difficult to assess and impossible to measure.  To overcome this, dose rate meters are set to measure an operational quantity known as the ambient dose equivalent rate.  The relationship between effective dose and ambient dose equivalent is quite complicated and is based on many assumptions, so we will not go into too much detail in this module.  However, it is sufficient to say that the measurement of ambient dose equivalent is set such that it gives a reasonable approximation of effective dose.

Dose rate meters measure the operational quantity of ambient dose equivalent rate.  For most applications, this measurement gives a good approximation of the effective dose rate to our bodies.

So, when we are measuring the ambient dose equivalent rate with our dose rate meters, we are actually getting a good approximation of the effective dose rate to our bodies.  For correctness, however, the term ‘dose’ is used throughout this module to mean ambient dose equivalent (where ambient dose equivalent is the operational quantity measured by area monitoring instruments).

One way of determining the total dose from a dose rate meter is by multiplying the exposure time by the measured dose rate (as discussed in Module 2.2 Protection from External Radiation Hazards).  However, some dose rate meters are able to sum (or integrate) the dose received over a given time and give a read out in dose units (Sv) rather than dose rate units.  These instruments are known as integrating or electronic dosimeters and they are particularly useful for monitoring personal dose rather than for area monitoring.  Integrating dosimeters are very light and portable as they usually contain a small G-M tube or a silicon PIN photodiode detector.  A typical integrating dosimeter is shown in Figure 3.

Figure 19

A Typical Integrating Dosimeter

An integrating dose rate meter should also be used to monitor the external hazard from some x-ray generators.  This is because the dose rate may be changing rapidly with time and many typical dose rate meters are unable to provide an accurate response to such radiation fields.

2.  Contamination monitors

Two types of contamination monitors are used in the workplace:

• Surface contamination monitors (for measuring contamination on surfaces); and

• Airborne contamination monitors (for measuring contamination in the air.)

Which ever type of monitor you are using, it is important to remember that no contamination monitor can measure all of the contamination present either on a surface or in air (i.e. no instrument is 100% efficient).  Each instrument will have different efficiencies (ranging from 0% to approximately 30%) for different radionuclides and an important result of this is that some monitors will not detect some radionuclides.

No radiation monitoring instrument is 100% efficient and some contamination monitors will not detect some radionuclides.

1.  Surface contamination monitors

Surface contamination monitors must be able to detect the presence of low levels of radioactive material on surfaces.  This is particularly important if the radionuclide is an alpha emitter, because even low concentrations may be able to enter the body and cause a significant internal dose.

Surface contamination monitor usually consist of a monitor connected to a detector (or probe) which is capable of measuring the different types of radiation.  A typical surface contamination monitor and two different types of probes are shown in Figure 4.

Figure 20

A Typical Surface Contamination Monitor and Two Probes

The radiation detected by a surface contamination monitor will be from the contamination present in the area under the probe.  This is the total contamination in that area and is made up of both the fixed contamination and the removable contamination.  Although the presence of any contamination in an area is not desirable, the amount of removable contamination is the most important to know as this is the radioactive material that could get into a persons body and lead to internal exposure.

The fixed contamination is only a radiological problem if it becomes removable or the dose rate from it is significant.  Section 5 of this module tells you how to determine the levels of fixed and removable surface contamination.

Most surface contamination monitors give a reading in counts per second (cps) or counts per minute (cpm).  However, it is more important to know what the surface contamination levels are in terms of activity per unit area (i.e. in units of becquerels per square centimetre (Bq cm^-2) or becquerels per square metre (Bq m^-2)) as this will enable you to compare your measured contamination levels with the limits given in national and international regulations.  So, if your monitor measures in cps or cpm, you will need to convert these values to Bq cm^-2 or Bq m^-2 (see Section 5 of this module).

Figure 5 shows the units used for measuring surface contamination.

Figure 21

Monitoring for Surface Contamination

Note that some surface contamination monitors display the reading directly in becquerels per square centimetre (Bq cm^-2).  However, when using this type of instrument, you must make sure that it is suitable for the radionuclide being measured (i.e that it has been calibrated for that radionuclide - see Section 1.3 of this module), otherwise the amount of contamination indicated on the monitor may be incorrect.

When using a surface contamination monitor which displays the reading in Bq cm^-2, you must make sure that the instrument has been calibrated for the radionuclide being measured.

2.  Airborne contamination monitors

Airborne contamination monitors must be able to detect and quantify radioactive contamination in air as this contamination has the potential to cause harm to humans.  Airborne contamination can be considered as both an internal and an external hazard due to the fact that not only can airborne contamination be breathed in (internal hazard), but also, submersion in a cloud of some radioactive gases can lead to a significant external dose (external hazard).

Monitoring for airborne contamination is more complex than monitoring for surface contamination as airborne contamination can exist in three forms, either as:

• particulates (i.e. consist of particles such as dust or smoke);

• gases (i.e. consist of gaseous radionuclides); or

• vapours (i.e. consist of small droplets of liquid radionuclides, or the gaseous form of a radionuclide which is usually a solid or liquid at room temperature).

For each type of airborne contamination, the monitoring process is slightly different.  However, in general, airborne contamination monitoring involves two stages.  Firstly, the air is either filtered or collected by using a pump known as an air sampler.  Then the count rate from the filter paper or air sample is measured using a radiation detector.  Although some air sampling equipment incorporates both the air sampler and the radiation detector (e.g. real time airborne contamination monitors), it is more usual to sample the air for a set period of time and then measure the count rate from the filter paper or air sample.  This count rate can then be converted into units of Bq m^-3 for comparison with national and international regulations.  Section 6 of this module details the methods of monitoring for each type of airborne contamination and also shows how the units may be converted.  Figure 6 shows the units used for measuring airborne contamination.

Figure 22

Monitoring for Airborne Contamination

Note that real time airborne contamination monitors give continuous readings of the airborne concentration in Bq cm^-3 or Bq m^-3.  However, they must be calibrated for the radionuclide being measured otherwise the readings will be incorrect.  Real time airborne contamination monitors can be particularly useful in areas where the levels of airborne contamination could present a problem as they can be set to alarm if air sampling results exceed pre-set values.

SELF-CHECK 1

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

1. a)Which part of a radiation monitoring instrument interacts with the incoming ionizing radiation?

b)Name two types of detector which can detect ionizing radiation.

2. What are the two types of instruments used for measuring ionizing radiation?

3. Fill in the gaps with a suitable word or phrase:

Dose rate meters usually give a direct reading in units of _________ ___ _____ (_____ ___).  They actually measure dose in terms of the _________ __________ ____ rate which is an operational quantity used to estimate the effective dose received by humans.

4. A dose rate monitor shows a dose rate of 2.5 mrem h^-1.  What is this dose rate in Sv h^-1?

5. Name the two types of contamination monitors used in the workplace.

6. Fill in the gaps with a suitable word or phrase.

Surface contamination monitors usually show readings in _______ ___ ________ (___) or counts per minute (cpm).  In order to compare these levels with national and international regulations, these readings need to be converted to units of _________ per _______ squared (__ ___).  Some surface contamination monitors give direct readings of surface contamination but when using this instrument you must make sure that the monitor is calibrated for the radionuclide being measured otherwise _________________________________________________________.

7.a)What are the three types of airborne contamination?

b)What units is airborne contamination measured in?

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

3.  Standards and Instrument Calibration

All radiation protection survey instruments must be calibrated against calibration sources representative of the sources the instrument will be used to measure.  This ensures that the instruments are capable of responding within acceptable levels of accuracy.  To ensure the national and international uniformity of response, these calibration sources must be traceable to a primary or secondary standard, and these standards are usually held at a national or international standards laboratory.  The primary standards used for the calibration of radiation protection monitoring equipment include exposure, dose rate, emission rate and activity.  Secondary standards are calibrated against primary standards and it is a requirement that national standards laboratories compare their standards against each other at regular intervals to ensure international uniformity.

Dose rate meter calibration is performed by determining the instrument response in a specific radiation field delivering a known dose rate at a particular set distance from the source.  The instrument controls are then either adjusted to read the desired dose rate or a calibration factor is determined and included in a calibration report.  As dose rate meters generally have different ranges (e.g. Sv, mSv, or Sv h^-1, mSv h^-1), response determination should be performed on all ranges.

A simplified diagram of dose meter calibration facility is shown in Figure 7.

Figure 23

A Typical Dose Rate Meter Calibration Facility

For instruments designed to measure surface contamination, there are a number of ways of calibrating the instruments.  The most common way is to determine the detector reading per unit surface activity.  Calibration factors will then be given to enable you to relate the detector reading to the actual the surface contamination in Bq cm^-2.

Instruments which give direct readings of airborne contamination levels in Bq m^-3 are calibrated by determining the detector reading per unit volume of airborne activity.  The instrument controls are then adjusted to read the desired airborne activity level.

It is important to remember that dose rate meters and contamination monitors are calibrated against specific sources (each with specific types and energies of radiation).  Although many dose rate meters are energy compensated for gamma radiation, and should have a relatively flat response over a wide range of gamma energies, in practice, the monitor may not be suitable to monitor gamma sources with energies different to the calibration source.  In addition, if the sources in the workplace are beta or x-ray emitters and the monitor was calibrated against a source with a single gamma energy, the monitor may not be suitable.  It is therefore important to ensure that the calibration is valid for the radionuclides being monitored.

When monitoring in the workplace, a wide range of radionuclides may be present.  You should therefore ensure that the monitoring equipment you use is calibrated for the sourcesbeing measured.

Sometimes it is impossible to find a monitor which has been calibrated for all the radionuclides you wish to measure.  In this case, when a mixture of radionuclides is present in a particular area, it is common to assume the contamination being measured is due to the radionuclide for which the instrument has the lowest response.  You will then overestimate the total contamination level, but you will be operating safely.  If the overestimated contamination is at a level that might represent a problem, then the actual radionuclide can be identified by estimating its’ energy by alpha, beta or gamma spectroscopy.

Calibration of instruments should be performed both before initial use as well as regularly during use.  The requirements for calibration frequency vary according to local regulations but usually one calibration per year is recommended for each instrument.

Details of the instrument calibration should be included in a calibration report.  This report should detail:

• any relevant calibration factors;

• the performance of the instrument over specific energy ranges;

• the conditions under which the instrument was calibrated;

• the response to the conditions for which it was designed; and  

• the response to all radiations and conditions for which it may be used in the field.

4.  Lower Limit of Detection

When performing radiological surveys, it is important to know what is the lowest dose rate or contamination detectable by the equipment.  Background radiation will always be detected by radiation detectors and this makes it very difficult to distinguish between natural background levels and other small amounts of radioactivity.

For laboratory counting equipment, an accurate lower limit of detection (LLD) can be calculated from the standard deviation (see Module 1.2 Basic Mathematics) of the background counts measured by the equipment.  For dose rate and contamination monitors this is not really possible and a more simple method is commonly used.  This simple method assumes that, because of the fluctuation in background and instrument reading, you can only be confident the monitor is detecting radiation in addition to background if the monitor reading is twice the background value.  (The background value should be measured in an area where there is no man-made radiation present.  This is not always possible and the background will be elevated.)  If the reading is not equal to or greater than twice the background, you do not know if there is any radioactivity present, all you can quote is that the value possibly present is less than the LLD.

For dose rate and contamination monitors, you can only be confident the monitor is detecting radiation in addition to background if the monitor reading is at least twice the background value.

The natural background dose rate level in most areas is not usually a radiological problem, so the dose rates measured are normally quoted directly from instruments.  However, when trying to measure contamination levels, the ambient background dose rate will produce a background reading on most contamination monitors and the actual contamination level is found by subtracting the background from the measured value (see Equation 1).

L = R – BG[1]

whereL = the actual contamination level

R = the reading on the contamination monitor

BG = the background reading

The LLD is an important factor you should take into account when monitoring low levels of contamination.  In some cases, contamination monitors cannot measure down to very low levels and to be able to quantify your results, you need to know what the lowest level of detection is.  Basically, the LLD (in cps or cpm) for a contamination monitor is twice background minus the normal background value.  This is because anything less than twice background cannot be considered valid, but the normal background value is attributed to ‘natural’ background levels rather than to the contamination being measured.  Hence the LLD of the monitor is the difference between these two values.  Obviously, twice background minus background equals background and hence the LLD (in cps or cpm) can be taken as the background level.  To convert this value to a meaningful reading in Bq cm^-2, we need to divide the background reading by the monitor’s calibration factor as shown in Equation 2.

LLD = 2BG – BG = BG[2]

CF           CF

whereLLD = the lower limit of detection in Bq cm^-2

BG = the background reading in cps

CF = the monitor’s calibration factor in cps per Bq cm^-2

An example of an LLD calculation is shown in Example 1.

EXAMPLE 1

Question

A contamination monitor has a background reading of 5 cps and the maximum reading seen on the monitor while monitoring a bench surface was 7cps.  The calibration factor for the monitor and radionuclide used in the area is 2 cps = 1 Bq cm^-2.  What is the lower limit of detection (LLD) of this monitor and how would we quote this reading?

Answer

The LLD can be calculated from Equation 2 as follows

LLD = BG = 5 = 2.5 Bq cm^-2.

CF    2

Hence, the LLD for this monitor is 2.5 Bq cm^-2.

From Equation 1:

L = 7 – 5 = 2 cps

As this contamination level is not equal to or greater than 2 times the background, we can therefore only quote that the contamination level less than the LLD i.e. the contamination level is less than 2.5 Bq cm^-2.

Note that, if the LLD of a monitoring system is below the level you are trying to measure (e.g. some regulations quote acceptable surface contamination levels of less than 1 Bq cm^-2), then this monitoring system is unsuitable for performing the measurement.

SELF-CHECK 2

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

1. Why is it important to calibrate radiation monitoring instruments?

2. Where must the calibration of instruments be traceable to?

3. How are dose rate meters calibrated?

4. What is the most common way of calibrating surface contamination monitors?

5. Why is it important to remember that dose rate and contamination monitors are calibrated against specific sources?

6. a)If contamination is found in an area containing a mixture of radionuclides, which radionuclide should you use to perform your calculations?

b)Why is this so?

7. In simple terms, how can you be sure that a reading on a dose rate or contamination monitor is a correct reading?

8. A contamination monitor has a background reading of 10 cps and the maximum reading seen on the monitor while monitoring a bench surface was 14cps.  The calibration factor for the monitor and radionuclide used in the area is 4 cps = 1 Bq cm^-2.  What is the lower limit of detection (LLD) of this monitor and how would we quote this reading?

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

2.  SELECTING APPROPRIATE INSTRUMENTATION

The ideal radiation monitoring instrument should be capable of monitoring all forms of ionizing radiation and should be portable, easy to use and directly indicate ambient dose equivalent rate or contamination levels.  In practice, it is not possible to design a single instrument to fulfil all of these requirements and hence, when performing a radiological survey, the instrument that best suits the needs of the job should be selected.

1.  Monitor Considerations

When selecting a suitable monitoring device for a specific radiological survey, the following items should be considered:

1. The type of survey you wish to perform (e.g. dose rate or contamination survey).

2. The type of radiation you wish to detect (e.g. alpha, beta, gamma, x-ray or neutron radiation).

3. The energy of the radiation you wish to detect (e.g. keV or MeV).

4. The required sensitivity of the detection instrument (e.g. Sv h^-1, mSv h^-1 or Sv h^-1, the maximum number of cps or Bq cm^-2).

Once these items have been considered, the selection of monitoring device can be narrowed down.

1. Type of survey

The type of instrumentation used to carry out a survey depends on the type of survey required.  When performing a detailed radiological survey, monitoring usually includes dose rate, surface and airborne contamination monitoring.  However, you may wish to measure only one of these things and it is important to know the type of survey required in order to select the correct instrumentation.

Remember that dose rates in the workplace are usually measured in units of Sv h^-1 or mSv h^-1.  Surface contamination levels are measured in counts per second (cps) (which may be converted to Bq cm^-2), and airborne contamination levels are usually measured in Bq m^-3.

2.  Type of radiation

As we have already seen in Module 1.5 Methods of Radiation Detection, different types of detector are sensitive to different types of radiation.  Therefore it is important to know which radiation you are looking for whenever a survey is performed.

When monitoring dose rates, remember that alpha radiation does not penetrate the dead layer of skin and therefore alpha external dose rates are not monitored.  In fact, dose rate meters are designed to monitor beta, gamma, x-ray and neutron radiations.

Dose rate meters can be designed to respond to one or more types of radiation but they may have different sensitivities to each type of radiation.  In mixed radiation fields, the type of radiation therefore needs to be identified if multiplying factors are to be used to determine the dose rate accurately.  Typically thin end window G-M tubes or ion chambers (see Figure 8) will measure photon and beta dose rates.  To tell the difference between these, a thin metal plate can be used to attenuate the beta particles.

Front View                                             Rear View

Figure 24

A Typical Ion Chamber Dose Rate Monitoring Instrument

Some neutron monitors also respond to gamma radiation and the neutron dose rate will then appear higher than it really is.  To overcome this the gamma response of the monitor and the gamma dose rate in the area should be known.  The neutron indication (which is really due to gamma radiation) can then be subtracted.  The gamma response can be determined at the routine calibration.

Contamination monitors may also respond to different types of radiation.  To help identify the radionuclides present and determine their significance, the type and quantity of each type of radiation needs to be determined.  This can be achieved by electronic discrimination of the probe signal or by using filters to attenuate one or more of the types of radiation.  The monitors that include discrimination are usually the most convenient to use.

You may find it useful to consider the following general points when you determine a suitable contamination probe:

40. Zinc sulphide scintillation probes and proportional counters can be used with alpha and beta discrimination.

41. Beta probes usually also respond to photon radiation, and a relatively low gamma background could lead to an overestimation of the beta contamination.

42. Alpha probes and phosphors do not usually respond to photon radiation.

43. G-M tubes respond to alpha, beta and photon radiation and cannot directly discriminate between them.

You will need to balance the advantages of the type of measurement you make with the disadvantages.  For example, if you wanted to monitor low levels of a particular radionuclide which emits alpha, beta and gamma radiations (such as natural uranium or thorium in equilibrium), as the alpha particles are likely to be attenuated, you could monitor the beta radiation.  However, if the gamma dose rate is high, the beta response could be a gross overestimate, so you will need to either monitor the alpha levels or use an absorber to distinguish the beta from the gamma reading.  A monitor calibrated in terms of gamma emissions from the source might also be used, but you may find the sensitivity is poor compared to the alpha or beta response.

3. Energy of radiation

The energy response of instruments measuring x-ray or gamma ray dose rates is important as dose rate survey meters are often calibrated using a particular source such as caesium-137 which has an effective photon energy of 0.66 MeV.  If the instrument is then used to measure radiation of different photon energy, it may overestimate or underestimate the dose rate.  In general, although the energy responses of the different types of GM detectors may be different, most instruments are now compensated to give a uniform response to x- and gamma radiation in the range 0.1 to 3 MeV.  Energy compensation of G-M detectors is usually achieved by surrounding most of the detector with a metal filter to selectively attenuate the lower energy radiations.

For contamination monitors the detection method and design of the detector determines the type and energy of the radiation that can be measured.  The specified response to the radionuclide of interest must therefore be checked.  Tritiated water and tritium (hydrogen-3) compounds are very difficult to monitor directly, so smears are commonly taken.  These smears are then monitored in a liquid organic scintillation counter.

4. Sensitivity of the instrument

The selection of dose rate meter depends on the sensitivity required.  For example, a simple ion chamber will measure a wide range of energies of beta, x- or gamma radiations but to monitor very low dose rates (tenths of microsieverts per hour), much larger chambers are required which are not as portable.

Dose rate units are usually displayed in Sv h^-1 or mSv h^-1, Gy h^-1 or mGy h^-1 and the range in dose rate depends on the application (see Table 1).

Table 13

The Range of Different Types of Monitoring

Type of Monitoring	Range
Occupational	0.1 Sv h-1 to a few mSv h-1
Environmental	nSv h-1 to hundreds of Sv h-1
Emergency	Sv h-1 to a few Sv h-1

Note that for emergency monitoring, the range required varies from Sv h^-1 to a few Sv h^-1.  For this type of monitoring, a special instrument known as a teletector is required.  This instrument incorporates a number of different range G-M tubes positioned at the end of an extending telescopic handle to maximise source to person distance (see Figure 9).

Figure 25

A Teletector

For contamination monitors, their sensitivity depends on the design of the detector.  The display range is usually in counts per second (cps) or directly in Bq cm^-2.  The range of a contamination monitor can typically be from zero to either a few hundred to a few thousand cps, (or the equivalent in the other units).  It is important that the maximum range you chose allows you to monitor the levels of contamination which might reasonably be expected in a particular area.  Otherwise, some instruments may indicate a full scale reading but you will not be able to tell the actual levels.  For radionuclide facilities, a maximum range of 5000 cps would be reasonable while in a counting laboratory a few hundred cps would be more appropriate.

As even low amounts of loose alpha contamination can present an internal radiation hazard, suitable alpha contamination monitors must be able to detect low levels of alpha contamination.

5.  Selection of monitoring device

Once the requirements listed in 2.1.1 to 2.1.4 above have been determined, the choice of instrument type is narrowed down.  Some of the more commonly used monitoring instruments are shown in Figure 10.

123 456

Dose Rate MetersContamination Monitors

Figure 26

Some Commonly Used Radiation Protection Instruments

Table 2 below shows the suitability of these detector types and others for dose rate and surface contamination surveys.

Table 14

The Suitability of Detector Types for Dose Rate and Surface Contamination Surveys

SELF-CHECK 3

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

1. What four aspects of a radiation survey must be considered before choosing an appropriate monitoring device?

2. A thin end window G-M tube can be used to monitor beta and gamma contamination.  How can we distinguish between these two types of contamination?

3. As the Radiation Safety Officer of a university, you are told that high dose rates are suspected in one of the research laboratories.  This particular laboratory uses only phosphorous-32 (a pure beta emitter).  Which type of monitoring instrument would be the most appropriate for performing a dose rate survey in this area?

4. You are responsible for performing random monitoring surveys in an area which deals with alpha emitters.  Which type of instrument would you use to directly monitor for alpha contamination?

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

2.  Design Considerations

Once the type of monitoring device has been narrowed down, there are a number of other design considerations which should be taken into account.  These are as follows:

1. The ease of use of the instrument (e.g. portability, design, simple display).

2. The suitability of the detector for the environment in which measurements are to be taken (e.g. effects of temperature and humidity, and whether it is intrinsically safe for hazardous atmospheres).

3. The location of the detector within the instrument (i.e. whether the detector is positioned in the body of the instrument or whether it is a separate probe on a cable).

4. The physical size of the detector for effectively measuring dose rates in narrow beams.

5. The response time for short exposures or high radiation fields.

6. The directional response of monitor.

7. The resistance to foldback in high radiation fields (for G-M detectors).

8. The presence of a speaker to help identify changing levels.

9. An illuminated scale for night work.

1.  Ease of use

Radiation instruments should be portable, have ergonomically designed displays/switches/buttons and be easy to decontaminate.  It should be easy to replace and check the batteries and read the display.

Neutron detectors usually incorporate a significant quantity of a hydrogenous material (like polyethylene) to thermalise the neutrons and this results in a heavy instrument that is not so portable.  With the present generation of neutron monitors their weight is a disadvantage that has to be accepted.

2.  Operating environment

The application and environment that the instrument is subjected to are also important factors in selection.  The effect of exposure to humidity, temperature and radiofrequency (RF) signals should be considered as these may affect the operation of the instrument.

Particular care must be taken in hazardous areas.  If there is a possibility of explosive gases in the area, as in the chemical or oil and gas industries, intrinsically safe monitors may need to be used.  Although intrinsically safe dose rate meters are available, at this time there are no intrinsically safe contamination monitors.

3.  Location of the detector

In some situations you may need to take a contamination or dose rate measurement in an area where normal access is difficult but which may be occupied in some other situation.  It is therefore helpful in this situation if the probe is on a cable rather than in the body of the detector so you can reach into the area and still read the monitor display.  This is common when monitoring the inside of vessels where contamination may be present or measuring dose rates in and around equipment where only hand access is possible.

4.  Physical size of the detector

To obtain an accurate dose rate measurement, the whole of the useful volume of the detector should be irradiated.  If this is not the case, the reading will be an underestimate.  If narrow radiation beams are to be monitored, a detector with a small cross section should be used.  In this case, the detector is usually on a cable rather than in the body of the monitor so you can locate the detector accurately at the position of the narrow beam.

5.  Response time

Where the radiation field is changing or is only present for a short period you will need to use a monitor with a fast response to be able to detect the radiation.  The disadvantage of a fast response is that the reading may fluctuate significantly and an accurate reading is difficult to obtain.  Manufacturers try to provide monitors which compromise between a fast response and a stable reading.

With some short exposures the response time of the monitor cannot be quick enough to measure dose rate.  In this case a total dose reading from an integrating dose meter over a number of exposures can be an alternative method.  From the total dose recorded, either the dose per exposure or the dose rate can be calculated by dividing the total dose measured by the number of exposures or the total time of all the exposures.

The speed of response of a monitor should increase with increasing dose rate.  You do not want to be waiting for a monitor to respond while you are in a high radiation field.

6.  Directional response

With contamination monitors it is easy to see that they will be more sensitive to radiation falling on the detector window.  With dose rate meters the manufacturers try to ensure the detector is sensitive to radiation arriving from any direction.  However in practice this is very difficult to achieve and they may provide a polar response graph which indicates the sensitivity for radiation from different angles.  Usually the monitor is most sensitive to radiation approaching from the front of the detector.

7.  Resistance to foldback

As you may remember from Module 1.5 Methods of Radiation Detection, G-M tubes can saturate in high dose rate fields and can then stop indicating the presence of radiation (this is known as fold-back).  You should therefore ensure that the monitor you use is protected from fold-back to ensure that the monitor continues to indicate a high value.

8.  Speaker

Speakers on both dose rate and contamination monitors are useful to quickly identify areas of high dose rate or contamination.  They are also helpful where you are using a probe on a cable and are watching where you are monitoring and do not want to keep looking at the instrument display.

9.  Illuminated scale

There may be occasions where measurements need to be taken outside at night or where visibility is poor.  A self-illuminating scale is a great help in this situation.  However, you should be careful as the light might significantly reduce the battery life of the monitor and spare batteries might be needed.

3.  Other Considerations

When purchasing monitoring equipment, two other aspects should be considered:

1.The cost of the instrument (e.g. initial purchase price and maintenance costs).

2.The availability of batteries and spare parts.

1.  Cost of the instrument

The initial purchase cost and the costs of routine maintenance and calibration must be considered when choosing a suitable instrument.  The vulnerability of a dose rate meter could determine the selection based on detector replacement alone.  Dose rate instruments usually incorporate the detector and display into one unit and occasionally this is completely sealed into a moulded case which makes repair very difficult.

2.  Availability of batteries and spare parts

In the past some monitors used batteries which were only available from the manufacturer and were expensive.  You should ensure the batteries are of a type commonly available.  You should also check the time it would take to have the monitor repaired by the manufacturer or supplier.  If spare parts are available, it is advisable for you to hold some in case the monitor needs minor repairs.

SELF-CHECK 4

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

1. Give three examples of design considerations which should be taken into account when selecting a monitoring instrument.

2. As a radiation safety officer at a radiopharmaceutical production facility, you are asked to assess the gamma dose rates in a small hand access area where maintenance work needs to be carried out.  You have a choice of an ion chamber instrument or a thin end window G-M probe (see Figure 10).  Which instrument would be most suitable for measuring the dose rates in this area?

3. A small radioactive materials storage facility has been constructed in a laboratory.  A suitable amount of lead shielding has been added but you have found that there is a short path (i.e. a narrow radiation beam) coming from the area where two lead sheets join.  What is the most important design consideration of a detector which could be used to accurately measure this dose rate?

4. You work in the x-ray department of a hospital.  The dose rates in this area change rapidly with time and many typical dose rate meters are unable to provide an accurate response to such radiation fields.  How could you estimate the average dose rate in this particular environment?

5. What are the two other aspects of monitoring instruments which need to be considered when purchasing equipment?

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

3.  PLANNING A RADIOLOGICAL SURVEY

There are a number of items to be addressed before a radiological survey can begin.  These are as follows:

1. Prepare radiological survey diagrams and forms for the area to be surveyed.

2. Refer to any previous surveys to estimate the dose rates and contamination levels expected in the area.

3. Perform appropriate pre-use response checks to ensure that the instrument is working correctly.

1.  Preparation of Forms and Diagrams

Before a radiological survey can begin, a diagram of the area to be surveyed should be drawn.  An example of a radiological survey diagram for a laboratory is shown in Figure 11.

Figure 27

Radiological Survey Diagram of a Laboratory

As shown in Figure 11, the locations where measurements are taken should be numbered.  These numbered areas can then be used as reference locations for future surveys of the same area and for detecting trends in measured values.  They can also be easily referred to on a radiological survey form as shown in Figure 12.  Note that dose rate measurements should be taken at contact distance and at 1 m to give an indication of the overall dose rates to people working in this area.  Contamination levels can be determined directly or by taking a smear of the area (see Section 5).

Other than providing a record of the radiological survey measurements, the survey form (Figure 12) also has space for the signatures and comments of the technician (i.e. the person performing the survey) and/or the radiation safety officer (RSO) (i.e. the person responsible for the radiological safety of the area).  The area supervisor (i.e. the person generally responsible for the area) should also sign the form to ensure that the comments are noted and that appropriate actions are being taken.

Figure 28

Radiological Survey Form

2.  Obtaining Information from Previous Surveys

Referring to previous surveys can give an idea of the dose rates and contamination levels expected in the area.  From these surveys, it may also be possible to decide whether time limitation or protective clothing is necessary.

3.  Pre-Use Response Checks

Before using any monitoring instrument it is necessary to ensure that it is operating correctly.  Hence, the following checks should be performed prior to use:

1. Visual check.

2. Battery check.

3. Calibration check.

4. Light sensitivity check.

5. Source check.

The frequency of these checks depends on how often the equipment is used.  However, the minimum recommended frequency is once per day of operation.  Note that if you have any doubts about the operation of a particular detector, it is better to use a different instrument and have the faulty one repaired as soon as possible.

1.  Visual check

The first check to be performed is a visual check of the instrument to ensure that it is not damaged in any way.  Visual checks should include an inspection of the outer casing of the instrument and also the probe.  If damage is visible (e.g a visible hole in the thin end window), the instrument should be repaired and recalibrated before use.

2.  Battery check

The first thing to check after an instrument is switched on is the battery.  For an instrument with an analogue display, the movement of a needle to a preset place on the dial usually indicates that the battery is in a good condition.  With electronic displays, the battery checks are automatically performed when the instrument is switched on and periodically whilst being used.  With this type of display, a low battery indicator is usually a feature of the instrument.  You should be aware of such design features which should be described in detail in the instrument manual.

If this check indicates that the battery is low, then the battery should be replaced before using the instrument.  In general, the instrument does not need to be recalibrated after a battery change unless it fails other operational checks.

3.  Calibration check

Before use, it is important to confirm that an instrument has been calibrated and that it is within its calibration dates.  This information is usually given in the instrument manual or on a calibration certificate.  The requirements for the frequency of calibration vary according to local regulations but usually one calibration per year is recommended for each instrument.  If an instrument is not within its calibration dates or it has been repaired for any reason, it should be recalibrated before it is used again.

4.  Light sensitivity check

Scintillation detectors operate by detecting bursts of light resulting from incident radiation interacting with the detector.  These detectors are sensitive to light and will give incorrect results if the thin window is punctured with a hole orthe construction is no longer light tight.  You can check this by holding the detector up to bright light and observing the response.  If the detector is sensitive to light, the response displayed will be very large and the instrument will need to be repaired and recalibrated. With some end window G-M detectors, strong sunlight can cause a current to flow across the tube and a signal to be recorded on the display.  This is usually when the metallic coating on the window peals off.

5.  Source check

Instrument performance may change depending on the environment and frequency of use.  For this reason, pre-use radioactive source checks should be used to detect and monitor any changes in performance quality.  For contamination probes, check sources giving greater than 100 cps with a surface area equal or larger than the probe size should be used for these checks.  A small source can also be used if it is placed in the same position relative to the probe face each time the check is carried out.  If alpha, beta and gamma contamination is to be monitored, then three independent sources are required.

To check dose rate instruments, the check source used should be of a similar energy and radiation type to that being monitored during surveys.  In addition, the radioactive source should give a dose rate of the order of 10 Sv h^-1 at 2 to 3 cm from the source.

The actual count rate or dose rate recorded by the instrument should be compared to the count rate or dose rate expected from the source.  This should be performed in a low background area and the background recorded and subtracted from instrument values.  If the two values (instrument and source) differ by more than  30% then the instrument should not be used.  Instead it should be repaired and recalibrated.

SELF-CHECK 5

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

1. What are the three items which need to be addressed before a radiation survey can begin?

2. Why is it important to perform pre-use response checks?

3. How often should pre-use response checks be performed?

4. Complete the following table with the names of the five pre-use instrumentation checks and what should be done if the instrument fails the test:

5. Why can it be useful to refer to previous surveys of the area?

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