TECHNOLOGY EXPERTS GROUP

BASIC PRINCIPLES OF

RADIATION PROTECTION

FOR RPO

Prepared by

Prof. Dr. M. FAROUK AHMAD

RIYADH

APR. 2006

2

FOREWORD

The use of man-made ionizing radiation and radioactive sources

are now a day widespread, and continue to increase around the world.

Nuclear techniques are in growing use in industry, agriculture, medicine,

well logging, and research benefiting the society as a whole. Irradiation

is used around the world to preserve foodstuffs. Sterilization techniques

have been used to eradicate diseases, and ionizing radiation are widely

used in diagnosis and therapy of different diseases. Industrial

radiography is widely used to examine welds and detect cracks and

microscopic bubbles in metallic pipes, tanks and other devices, and help

prevent the failure of engineered structures.

It has been recognized that exposure to a an acute dose of ionizing

radiation causes clinical damage to the tissues of the human body. In

addition, long term studies of populations exposed to ionizing radiation

have demonstrated that this exposure has a potential for the delayed

induction of malignancies. Due to these risks all activities involving

radiation exposure shall be subjected to certain national and international

safety standards, in order to protect radiation workers, general public and

environment from exposure to ionizing radiation.

One of the requirement of the national and international safety

standards is that any installation, that is acquiring any of the radiation

sources shall appoint a radiation protection officer, RPO, (or officers), to

oversee the application of the requirements of the radiation protection

and safety of radiation sources. According to the Saudi national and

international regulations, this individual shall be technically competent

in radiation protection scientific and organizational matters, relevant for

a given type of practice. In Accordance with Saudi national regulations

shall be licensed by the national regulatory authority through passing a

qualification exam, which is held periodically by this authority.

For successfully passing this qualification exam, one should study

different scientific and organizational topics, which are existing in

different English books, and are specialized very deep in the subjects of

interest. It may be very difficult for individuals non specialized in

radiation physics to follow this subjects.

3

For this reason this booklet is prepared, and will be issued, by the

technology experts group, to cover the fundamentals and all scientific

and organizational topics that are necessary for any radiation protection

officer to be qualified as a RPO. Together with the included topics in

this booklet the practical lessons are essential part of the qualification of

the RPO. This practice in the different relevant fields may be gained

easily through these practical lessons.

We hope that the booklet will be helpful in acquiring the necessary

knowledge in the field.

Technology experts Group

and the author

4

PATRT 1

SCIENTIFIC AND TECHNOLOGICAL ASPECTS

OF RADIATION PROTECTION

5

CONTENTS

Part 1: Scientific and technical aspects of radiation protection.

Chapter 1: Radioactivity and radioactive decay.

1-1 Some properties of atomic nuclei.

1-2 Some properties of alpha decay and alpha particles.

1-3 Some properties of beta decay and beta particles.

1-4 Some properties of gamma disintegration.

1-5 The x-rays.

1-6 The neutrons and their sources.

1-7 Calculation of the source activity

1-8 The units of activity.

1-9 The physical half-life time.

1-10 The biological and effective half-life times.

1-11 The radioactive decay law..

1-12 The relation between the decay constant and the half-life time.

1-13 Some important multipliers.

Chapter 2: Interaction of radiation with matter.

2-1 Introduction.

2-2 Interaction of heavy charged particles with matter.

2-3 Interaction of beta particles with matter.

2-4 Interaction of x and gamma radiation with matter.

2-5 Interaction of neutrons with matter.

Chapter 3: Radiation detectors, survey meters and monitors.

3-1 General.

3-2 The gas detectors.

3-3 The scintillation detectors.

3-4 The semi-conductor detectors.

3-5 The survey meters.

3-6 The contamination monitors.

3-7 Devices for personal dosimetry.

Chapter 4: Some radiation measurement techniques and statistical

fluctuations.

4-1 Introduction.

4-2 The solid angle.

4-3 The detector intrinsic efficiency.

6

4-4 Relation between the counting rate and source activity.

4-5 Other factors affecting the measurements.

4-6 Dead time correction.

4-7 The statistical fluctuation of radiation measurements.

Chapter 5: Dosimetry quantities and their units.

5-1 The exposure.

5-2 The absorbed dose.

5-3 The equivalence between the Roentgen, the rad and Gray.

5-4 The Kerma

5-5 The radiation weighting factor.

5-6 The equivalent dose.

5-7 The tissue weighting factor.

5-8 The effective dose.

5-9 The committed equivalent or effective dose.

Chapter 6: Biological effects of radiation.

6-1 Direct and indirect action of ionizing radiation on cell.

6-2 Radiation effects.

6-3 Deterministic and stochastic effects.

6-4 Acute deterministic effects.

6-5 The stochastic effects.

6-6 Hereditary effects.

Chapter 7: Dose calculation.

7-1 Dose calculation from point sources.

7-2 Dose calculation for beta emitters.

7-3 Dose calculation from external gamma sources.

7-4 Dose calculation from neutron sources.

7-5 The inverse square low for external exposure

7-6 Dose calculation from internal exposure.

7-7 The annual limit on intake.

7-8 The derived air concentration.

Chapter 8: Radiation shielding.

8-1 Shielding of sources of alpha particles.

8-2 Shielding of sources of beta particles.

8-3 Shielding of x and gamma ray sources.

8-4 Shielding of the neutron sources.

7

Part 2: Organizational aspects of radiation protection.

General framework and requirements for radiation protection.

1- Introduction.

2- Administrative requirements.

3- Management requirement for radiation protection.

4 - The principle requirements.

5- Verification of safety.

6- Condition of service.

Responsibilities of parties.

1- Responsible parties for radiation protection.

2- Responsibilities of the licensee.

3- Cooperation between licensees and employers.

National (SA) dose limits.

1- The terms limit and level.

2- Radiation exposures.

3- The occupational dose limits.

4- The dose limits for general public.

5- The dose limits for medical exposures.

6- The dose limits for emergency exposures.

The radiation Protection Program (RPP).

1- Introduction.

2- The structure of the RPP.

The safe transport of radioactive material.

1- Introduction.

2- Definitions.

3- General provisions.

4- Determination of the transport index.

5- Categories of packages.

6- Marking and labeling.

7. Storage in transit.

8

CHAPTER 1

RADIOACTIVITY AND RADIOACTIVE DECAY

1-1 Some properties of the atomic nuclei:

- Any atom is composed of the atomic nucleus, around which

electrons are orbiting in elliptical shells.

- The radius of the atom is in the order of 10-10

m, while the radius

of the nucleus is in the order of 10-15

m, so that the volume of the

nucleus is smaller than that of atom by about thousand trillions times

(trillion = 1012

). Due to these dimensions, the atom is similar to the solar

system, with its inter- planetary distances.

- Any atomic nucleus consists of nucleons, which are protons or

neutrons. The proton mass is, approximately, higher than that of the

electron by about 1836 times, while the neutron mass is higher by about

1838 times. So, the neutron and the proton may be considered as

particles with the same mass. From these data the atomic mass is

concentrated in the atomic nucleus, and the nuclear density is,

approximately, constant and equals 1017

kg/m3 (about 100 millions

ton/cm3).

- The charge of the proton equals to the electron charge in

magnitude (1.6x10-19

Coulomb), but it is positive in sign, while the

neutron is neutral (e.g. its total charge equals zero). So, in a neutral atom

the number of the protons in the nucleus equals the number of the orbital

electrons.

- The number of the protons in a nucleus is called its atomic

number Z, while the total number of protons and neutrons, in it, is

called the mass number A. So the number of neutrons N in a nucleus is

N = A – Z. Symbolically, any atom is represented by the first letter

written in capital, or by the first one in capital and other one written in

small. The atomic number is written in the lower left corner, while the

mass number is written in the upper left one. Example of that is C12

6 (or

carbon-12), Cl35

17 (or chlorine-35), Cr51

23 (chrome-51) and Cd114

48 (or

cadmium-114).

- The nucleus of any element is composed of the same number of

protons Z, but it may have different numbers of neutrons N. these

9

different forms of the same element are called isotopes of the element.

For example, hydrogen exists in three forms (the nucleus of each

contains one proton), H1

1 without any neutron, H2

1 (or deuterium) with

one neutron and, H3

1 (or tritium) with two neutrons. The isotopes of the

element are characterized by the same chemical properties while they

have different physical properties. Some Elements have more than 40

isotopes.

- Some nuclides are stable, while some others are unstable and

they may, spontaneously, decay to daughter nuclides through the

emission of alpha or beta particle, or may disintegrate through the

emission of gamma radiation. These nuclides are called radio-nuclides

and there atoms are called radio-active isotopes. So, there are three types

of the radioactive decay, which are:

a) alpha decay (α decay)

b) beta decay (β decay), and

c) gamma disintegration (γ disintegration)

1-2 Some properties of α-decay and α-particles:

- In α decay of a nucleus, an alpha particle (α), which is the

nucleus of a helium-4 atom ( He4

2 ), is emitted. This particle is composed

of 2 protons and 2 neutrons. So, in an α decay of a parent radionuclide

the mass number of the daughter nuclide is reduced by 4 while the

atomic number is reduced by 2. An example of alpha decay is the decay

of uranium-238 to thorium-234 with the emission of an alpha particle α,

which is symbolically represented as:

U238

92 Th234

90 + He4

2

Another example is the decay of polonium ( Po210

84 ) to the stable

lead-206 ( Pb206

82 ) which is symbolically represented as:

Po210

84 Pb206

82 + α

- Alpha particles emitted from a certain radionuclide are

characterized by, so called, discrete spectrum. This means that all alpha

particles emitted from that radionuclide will have the same energy value

or separated but fixed values. So, by measuring the energy value or

values of α particles the radionuclide can be easily identified. In other

1

0

words, it is known that U238

92(for example) emits α particles with two

energy values which are 4.196 and 4.149 MeV. So, if these two energy

values for any alpha emitter are detected, then it mean that this emitter is

U238

92.

1-3 Some properties of β-decay and β –particles:

- There are three types of beta decay, which are:

1-3-1 Electron or β-negative decay:

- in this type of β decay one of the neutrons n of the parent nucleus

decays, spontaneously, to a proton p, negatron β- (which is a β-negative

particle i.e. electron) and a third particle, named anti-neutrino υ-. This is

represented symbolically as;

n p + β- + υ

-

- One example of β- (or electron decay) is the decay of

Co60

27 (Cobalt-60) to Ni60

28 (Nickel-60) with the emission of β- particle and

anti-neutrino υ-(see fig. 1-1), which is expressed symbolically as:

Co60

27 Ni60

28 + β- + υ

-

- Other example is the decay of cesium-137 to barium-137 with the

emission of the same two particles (see fig. 1-2). This is expressed as:

Cs137

55 Ba137

56 + β- + υ

-

- It should be mentioned that the decay energy which is a fixed

amount for each parent radionuclide to decays to a daughter one is

distributed randomly between the two emitted particles, β- and υ

-. In

some decays of the parent radionuclide the majority of the fixed decay

energy is acquired by beta particle, and the remaining small amount of

energy is acquired by the anti-neutrino. In other decays of the same

parent radionuclide the beta particles acquire a medium or a small

amount of the decay energy, and hence the anti-neutrino will get a

medium or a large amount of the decay energy. That is the reason of

emission of beta particles from the same radionuclide with energies

varying from zero up to the maximum decay energy. This is expressed,

in other words, in that the beta spectrum of any beta emitter is a

1

1

continuous one for different types of beta decay, and by studying beta

spectra it is impossible to identify the beta-emitting radionuclide.

- In beta-negative decay the mass number A of both parent and

daughter radio-nuclides remains constant and does not change, while the

atomic number Z of the daughter nuclide is increased by one with

respect to that of the parent one, since a neutron is converted into a

proton in the nucleus.

1-3-2 Positron or beta positive decay:

- In this type of β decay one of the protons of the parent nucleus

decays spontaneously to a neutron, β+ (which is a β-positive particle i.e.

positron) and a third particle, named neutrino υ. This is represented

symbolically as;

p n + β+ + υ

- One example of β+ (or positron decay) is the decay of Na-22

(Sodium-22) to Ne-22 (Neon-22) with the emission of β+ particle and

neutrino υ (see fig. 1-1), which is expressed symbolically as:

Na22

11 Ne22

10 + β+ + υ

- In beta-positive decay the mass number A of both the parent and

daughter radio-nuclides remains constant and does not change, while the

atomic number Z of the daughter nuclide is decreased by one with

respect to that of the parent one, since one proton of the parent nucleus is

converted into a neutron.

1-3-3 The electron capture:

- In this type of β decay one of the protons of the parent nucleus

captures an orbital electron from the shells, which are very close to the

nucleus, forming a neutron and a neutrino υ is emitted during this

process. This is represented symbolically as;

p + e- n + υ

- One example of the electron capture is the capture of an orbital

electron by Na-22 (Sodium-22) nucleus to form a Ne-22 (Neon-22)

nucleus with the emission of a neutrino υ. This is expressed symbolically

as:

1

2

e- +

Na2211 Ne22

10 + υ

- In the electron capture no beta particle is emitted, but the only

emitted particle is the neutrino. Moreover the mass number A of both the

parent and daughter nuclides remains constant and does not change, as in

all other types of beta decay, while the atomic number Z of the daughter

nuclide is decreased by one with respect to that of the parent one, since a

proton is converted into a neutron, by the analogy to the beta positive

decay.

1-4 Some properties of gamma disintegration:

- If an atomic nucleus is formed in, so called, excited energy state

(i.e. in a state with excess energy) it may disintegrate to a state with a

lower excitation energy or to the so called, the ground state (i.e. to the

state with zero excitation energy). This disintegration is accompanied

with the emission of a gamma (γ) photon, that carries an amount of

energy equal to the difference between the excitation energies of the

initial and final states. So, the energy Eγ of the emitted γ photon is given

as:

Eγ = Ei - Ef

where Ei and Ef are the excitation energies of the initial and final states

of the gamma emitting nucleus, respectively.

- Each γ photon is an electromagnetic wave (with zero rest mass)

with an ultra-high frequency f of a given value, which is, in its turn, a

characteristic value for this disintegration.

- An example of gamma disintegration is the disintegration of

*60

28 Ni nucleus, which is formed in an excited state, as a result of beta

decay of the Co60

27 , with an excitation energy equal to 2505 KeV, and then

it disintegrates, promptly, to a lower excited state with an excitation

energy equal to 1332 KeV, which, in its turn, disintegrates promptly to

the ground state with zero excitation energy. This means that the *60

28 Ni

emits two γ photons, one with energy Eγ1 = 2505 – 1332 =1173 KeV,

and the second with energy Eγ2 = 1332 – 0 = 1332 KeV. These two

gamma ray photons are characteristic lines (i.e energies) for the gamma

disintegration of *60

28 Ni , and hence for the decay of the Co60

27 to *60

28 Ni .

1

3

So, the detection of two gamma ray lines with energies 1173 and 1332

KeV is an indication that the original radio-nuclide is Co60

27 .

Fig (1-1): β decay of Co-60 and gamma disintegration of Nickel-60

- Other example of gamma disintegration is the disintegration of

*137

56 Ba nucleus, which is formed in an excited state, as a result of beta

decay of the Cs137

55 , with an excitation energy equal to 662 KeV, and then

it disintegrates, promptly to the ground state with zero excitation energy.

This means that the *137

56 Ba nucleus emits one γ photon with energy Eγ

= 662 – 0 = 662 KeV. This gamma ray photon is a characteristic line for

the gamma disintegration of *137

56 Ba , and hence for the decay of the

Cs137

55 to *137

56 Ba . So, the detection of one gamma ray line with energy

662 KeV is an indication that the original radio-nuclide is Cs137

55 .

- Gamma ray photons emitted from a certain radionuclide are

characterized by, so called, discrete spectrum. This means that all

photons emitted from that radionuclide will have the same energy value,

2505 KeV

1332 KeV

1173 KeV γ photon

1332 Kev γ photon

Ni60

28

Co60

27

1

4

as in the case of Ba-137, where the energy of the emitted photons is 662 KeV, or

separated but fixed values, as in the case of Co-60 where photons are

emitted with two discrete energies 1173 and 1332 KeV. So, by measuring

the energy value or values of gamma rays the radionuclide can be easily

identified. In other words, if photons with energy equal to 662 KeV (for

example) are detected, then this means that this emitter is Cs-137, and if

photons with energies 1173 and 1332 KeV are detected it means that the

emitter is Co-60

Fig (1-2): β decay of Cs-137 and gamma disintegration of Barium-137

- It should be noticed, that in gamma disintegration, neither the

atomic number Z nor the mass number A change. This is

expressed symbolically by the following gamma

disintegration:

CoCo 60

28

60

28 *

BaBa 137

56

137

56 *

- It should be also mentioned, that gamma emitters can be obtained

as a result of alpha or beta decays, when the daughter nuclei are formed

in their excited states. Gamma emitters may be obtained, too, by forming

excited states of nuclides during different nuclear reactions. If the half-

life time of the excited states is extremely short then the gamma

Cs137

55

*137

56 Ba

Ba137

56

662 KeV

KeV

lineγ662 KeV oβ

β1

0 KeV

1

5

disintegration will be prompt. In case, if the half-life time of the excited

states is long, then this state is called metastable, and the gamma

disintegration occurs during relatively long time. An example of the

metastable radio-nuclides, which is widely used in medicine as a gamma

emitter, is technicium-99 (Tc-99).

1-5 The x-rays:

- The x-rays are electromagnetic radiation, emitted either: a) as a

result of the interaction of the charged particles (mainly light particles

such as the electrons) with the negative orbital electrons or the positive

atomic nuclei or, b) as a result of the transfer of an orbital electron from

an orbit with higher energy to another one with lower energy. So, based

on the origin of x-ray there are two types which are bremstrahlung

and characteristic x-rays. The frequencies of these rays lay in the

region from about 1x1017

up to about 1x1022

Hz and even higher. So, the

x and gamma radiation are widely overlapping with respect to their

energies.

- An example of the bremstrahlung x-rays, is the x-rays which

are emitted from x-ray tubes as a result of acceleration of the electrons

by a voltage difference, and then braking these electrons by high Z

elements (e.g. in the electric field of the orbital electrons and nuclei).

These bremstrahlung rays are characterized by a continuous energy

spectrum, (e.g energies of the photons may vary from zero up to the

maximum energy of the accelerated electrons). With some

approximation, the average energy of the x-ray photons may be

considered equal 0ne third of the energy of the accelerated electrons.

- An example of the characteristic x-rays, is these x-rays which

are emitted as a result of the transfer of an electron from an orbit with

higher energy to another one with lower energy, when there is an

electron vacancy in the lower shell. Since electronic orbits have definite

discrete energy values for each element, there will be a characteristic x-

ray discrete spectrum for each element. This means that x-ray will be

emitted from all atoms of same element with the same definite energy

values, which are characteristic values for this element.

1-6 The neutrons and their sources:

- As it has been mentioned, the neutron is a neutral particle (e.g.

with total charge equal zero and with rest mass, very slightly, higher

1

6

than that of the proton. There are no naturally occurring radionuclides

that can emit neutrons. There is only one artificial (man-made)

radionuclide which can partially decay through the emission of a neutron

or with the emission of alpha particles. This is the californium-252 (Cf-

252) which is an alpha and neutron emitter with a half-life time of 2.64

years

- The most commonly used neutron sources in industrial and other

applications are: the americium-beryllium (Am242-Be9) source, the

californium- 252 and the neutron generators. The nuclear reactors are

used as a very powerful neutron sources with a neutron density ranging

from 1013

up to 1018

per cm3. These reactors are used for energy

production, as well as for thermal neutron irradiation for production of

different artificial radioisotopes.

- Neutrons emitted from all neutron sources, generators and even

reactors are fast neutrons, and their energies varies about zero up to

about 14 MeV.

1-6-1 The americium-beryllium neutron sources:

- The (Am242-Be9) neutron source is made by mixing a certain

amount of a very fine powder of americium-242 with a certain

weight of a very fine powder of beryllium-9. The Am-242 is a

source of alpha particle, which interacts with a beryllium

nucleus and produces a neutron, in accordance with the

following nuclear reaction:

He4

2 + Be9

4 C12

6 + n1

0

- This reaction is expressed in other form of writing as (, n)

reaction on beryllium, where denotes the projectile alpha

particle and n denotes the resultant neutron emitted in the

reaction, while beryllium denotes the target atom. Activity of

one Curie (1Ci) of Am-242 with about one gram of Be-9

produces a neutron source, with a neutron yield of about,

2.2x106 neutrons / second. Earlier, neutron sources were

made of radium-226 or Po-210, (as alpha emitters) with

beryllium-9. However, but the production of such sources has

been stopped due to the explosion hazards of Ra-226 or

relatively short half life time of Po-210. In all alpha beryllium

1

7

neutron sources, fast neutrons are emitted with energies

varying between zero and about 10 MeV

1-6-2 The californium-252:

The californium-252, which is an isotopic neutron sources, is

produced in nuclear reactors. 1 microgram (1 μg) of Cf-252 produces

about 2.3x106 fast neutrons per second. Neutron sources with different

yields ((up to more than 10 milligrams, e.g. 2.3x1010

neutrons/second)

are available in the market. Energies of the emitted neutrons from this

source vary from about zero up to more than 8 MeV.

1-6-3 The Photo-neutron source:

- In this type of neutron sources a gamma source which can emit

photons with energy higher than 1.67 MeV is used to interact with

beryllium-9 and split it to two alpha particles and a neutron according to

the following photonuclear reaction:

γ + Be9

4 2 He4

2 + n1

0

- The most commonly used gamma emitter in the photo-neutron

sources is sodium-24 (Na-24), which emits gamma photons with energy

of 2.76 MeV. The fast neutrons emitted from this source are

characterized by a mono-energetic value (e.g. all emitted neutrons

will have the same energy) instead of the continuous energy

spectrum which is obtained from all alpha-beryllium sources.

1-6-2 The neutron generators:

- These devices are small accelerators in which deuterons (denoted

as d, H2

1 or D2

1 , which is an isotope of the hydrogen) are accelerated

using a potential difference of about 150 Kilo- Volt (KV), to gain energy

of about 150 KeV, and then they collide a tritium (denoted as H3

1 or T3

1 )

target (tritium is another isotope of the hydrogen) to yield an alpha

particle and fast neutrons in accordance with the following nuclear

reaction:

D2

1 + T3

1 He4

2 + n1

0

which is known as (deuteron, neutron) reaction on tritium, and

which can be written as (d, n) reaction on tritium.

1

8

- The neutrons are emitted from this reaction with a fixed energy

value of 14.1 MeV. Neutron generators of this type are produced with

different neutron yields, varying from about 106 up to 10

12

neutrons/second.

1-6-3 The nuclear reactors:

The nuclear reactor is a facility in which neutrons are obtained as

a result of the fission of a fissile material, such as U-235 or Pu-239, in

sustained chain reactions. The emitted neutrons from the nuclear fission

are fast. However, they are moderated (slowed down) to thermal

neutrons by a moderators which ,usually, is light or heavy water or

graphite. Most of the reactors used for different applications are operated

with thermal neutrons. The neutron density in the reactor core varies

from about 1013

up to 1018

neutrons/cm3, depending on the reactor

power.

1-7 Calculation of the source activity A:

- The activity A (in decay per second) of a certain radioactive

source or sample is defined as the number of decays (or disintegrations)

that occur in this source or sample in a unit of time. In the SI system

units the time is expressed in seconds (s). If the source contains at a

certain moment N radioactive atoms, and if the probability for a single

atom of this type, to decay per second is λ (1/s) then the activity of this

source is equal λ N decays/second: e.g:

A = λ N (1-1)

1-8 The specific activity:

- The specific activity is the activity of a unit of mass, volume,

area or length. It represents the amount of activity existing in any of

these massive, volumetric, surface or line samples or species.

1-9 The decay (or disintegration) constant λ:

The probability for a single atom of a certain radionuclide to

decay per second is called the decay constant λ of this nuclide and its

unit in SI system is (1/s) i,e s-1

.

1

9

1-10 The units of Activity, The Becquerel and the Curie:

- In the SI system of units the activity A is measured in Becquerel

(Bq), which is one decay (disintegration) per second. So, in a sample

with 15 Bq activity, 15 decays occur per second from the parent nuclide

to the daughter one.

- In the old system of units source activity was expressed in

Curie (Ci). One Ci was defined as the activity of one gram of pure

radium-226. Later, it has been determined that one Ci is equal to 3.7 x

1010

decays/second. So, the relation between the Ci and the Bq is:

1 Ci = 3.7 x 1010

Bq

- The SI units of the specific activity are:

* Bq/Kg for massive species, such as food, soil and other

samples

* Bq/m3 for volumetric samples, such as air, water and

other samples

* Bq/m2 for surface samples such as surface contamination.

* Bq/m for line samples such as long pipes or rods.

- In other systems of units the specific activity may be expressed

in Curies/gm, Bq/liter, Ci/m3, Ci/cm

2, Ci/cm, or many other units. One

should be able to transfer from these units to those of the SI system and

vice verse.

1-11 The physical half-life time T1/2:

- The physical half-life time Tp1/2 of a radio-nuclide, or simply the

half-life time T1/2 is defined as the time period during which one half of

the total number of that nuclide decays (disintegrate) and the other half

remains without decay (disintegration). So, if (for example) the T1/2 of a

certain radio-nuclide is 5.27 years, and if at a certain moment we have a

sample of that nuclide containing 4000 radioactive atoms, then during

5.27 years 2000 atoms decay and the other 2000 remain without decay.

During the second 5.27 years one half of the remaining atoms decays

(e.g 1000 atoms decay and the other 1000 remain without decay).

During the third 5.27 years 500 atoms decay and the other 500 remain

without decay etc.

2

0

1-12 The biological and effective half-life times:

- When a human being is ingesting or inhaling, any radio-active

isotope (or radio-nuclide) by injection or through a wound, then the

amount of the radio-nuclide in the body will be reduced as a function of

time due to two different effects, which are:

a) The physical decay of the radionuclide, with the physical

half-life time T1/2, which is not affected by any physical,

chemical or biological factors.

b) The different biological excretion processes, such as urine

and other excreta, with biological have life-time Tb1/2

- The biological half-life time Tb1/2 is defined as the time period

during which one half of the total number of that ingested, inhaled or

injected radio-nuclide will be excreted out from the human body,

through all excretion processes, and the other half remains inside the

body. It should be mentioned that although the Tb1/2 is considered

constant, it may vary in limited way, from man to other, depending on

the human dietary food habits.

- The effective half-life time Te1/2 is defined as the time period

during which one half of the total number of that ingested, inhaled or

injected radio-nuclide will be decayed or excreted out from the human

body, through the physical decay process and all excretion processes,

and the other half will remain inside the body without decay. The

effective half-life time Te1/2 is related with both the physical half-life

time Tp1/2 and the biological half-life time Tb1/2 by the following simple

relation:

(1/ Te1/2) = (1/Tp1/2) + (1/Tb1/2) (1-2)

1-13 The radioactive decay law:

- This law relates the number of remaining atom without decay N

with respect to its initial number N0 as a function of the time t. This

relation is expressed as:

N = N0 e – λ t

(1-3)

- The same law is used to express the exponential decrease of a

sample activity A with respect to its reference activity A0 at a certain

2

1

reference moment t = 0, as a function of time t. It is expressed in the

following form:

A = A0 e – λ t

(1-4)

1-14 The relation between decay constant λ and the half- life time

T1/2:

- Using the radioactive decay law and the definition of the half-life

time T1/2 it is easy to show that the decay constant λ is related with the

half-life time T1/2 by the following simple relation:

λ = ln2 / T1/2 or

λ = 0.693 / T1/2 (1-5)

- The biological decay constant λb is related with the biological

half-life time Tb1/2 with a relation of the similar form e.g:

λb = 0.693 / Tb1/2

and the effective decay constant λe is related with the effective

half-life time Tb1/2 with a relation of the form:

λe = 0.693 / Te1/2

- The effective decay constant λe is related with the effective the

physical decay constant and the biological decay constant as:

λe = λp + λb (1-6)

2

2

1-15 Some important multipliers

Subscripts Notation The multiplier

1 deci 1 d 1 x 10-1

1centi 1 c 1 x 10-2

1 milli 1 m 1 x 10-3

1 micro 1 μ 1 x 10-6

1 nano 1 n 1 x 10-9

1 pico 1 p 1 x 10-12

1 femto 1 f 1 x 10-15

Superscripts

1 Deco 1 D 1 x 101

1 Hekto 1 H 1 x 102

1 Kilo 1 K 1 x 103

1 Mega 1 M 1 x 106

1 Gega 1 G 1 x 109

1 Tera 1 T 1 x 1012

1 Exa 1 E 1 x 1015

2

3

2

4

CHAPTER 2

INTERACTION OF RADIATION WITH MATTER

2-1 Introduction

From the view point of interaction between particles or radiations

and matter, particles and radiations are divided into four different

groups. These are:

a- Heavy charged particles, such as alpha particles, deuterons, and

protons.

b- Light charged particles, such as beta particles (which are

electrons and positrons).

c- Electromagnetic radiations, such as x-rays and gamma radiations.

d- neutral particles such as neutrons.

2-2 Interaction of heavy charged particles, with matter:

- When a parallel beam of heavy charged particles, such as α

(alpha) particles or protons is incident on a matter, these particles

interact, mainly, with the orbital electrons of the atoms, which form this

matter, through the Coulomb forces that arise between the charge of the

incident particle and the orbital electrons. The interaction between the

incident particles and the atomic nuclei of the matter is too limited, from

the point of view of radiation protection. This Coulomb interaction

(due to Coulomb force between the incident charged particle and the

orbital electrons) results in transferring a portion of the energy from the

incident particle to the orbital electrons. If the transferred energy is

relatively low (within some eV), then the affected electron can be

removed from its orbit to another one in the same atom with higher

orbital energy, in a process called "excitation". If the transferred

energy is relatively large, then the affected electron will be kicked

out from its mother atom, in a process called "ionization", where

the electron (with its negative charge) becomes free and the atom

becomes ionized with positive charge, e.g. positive ion. In other words

the energy transfer will lead to formation of the so called electron-ion

pair. In case, if the transferred energy is larger enough (within some

hundreds of eV) then the kicked electron, in its turn, may ionize a

2

5

neutral atom forming a new electron-ion pair or pairs. In this case

electrons are called delta () electrons. The main properties of the

interaction between heavy charged particles and matter can be

summarized in the following:

- The main processes by which alpha particles with relatively low

energies (5-10 MeV) transfer their energy to the matter is the ionization

and excitation.

- The track of any heavy charged particle in the matter is a straight

line (due to the large mass of the incident particle with respect to the

electron mass).

- The energy is transferred from the incident heavy charged

particle to the electrons in relatively very small portions. This means

that the energy of the incident heavy charged particle is reduced

gradually as it penetrates through the matter. At the end of the track, the

alpha particle will capture two electrons from the neighbor atoms

forming an inert atom of helium-4.

- The average energy w, which is required to form one

electron-ion pair in air or human tissue is about 34 eV, so that, the

average number of electron-ion pairs formed in the whole range of 5 MeV

alpha particles is about 150000 pairs.

- The delta electrons represent about 70 % of the total number

of free electrons, while the primary electrons represent about 30 %

only.

- Different particles with the same incident energy will have

slightly different rang inside the matter. This effect is called :stragling".

- the range of 5 MeV alpha particles is about 35- 40 mm in air at

standard temperature and pressure, and about 40 micrometers in water or

human tissues.

- The specific ionization s of alpha particles with about 5 MeV

energy in air, which is defined as the number of electron - ion pairs,

formed in 1 mm of their track, varies from about 2000 pairs/mm at the

beginning of the track to more than 6000 pairs/mm at the end of the track.

Fig. (2-1) shows the variation of s as a function of penetration distance

in the matter.

- The stopping power (dE/dx) of alpha particles in a matter, which

is defined as the amount of energy transferred per unit length of the track

2

6

is given as the product of the energy w needed to form one electron- ion

pair by the specific ionization s, e.g:

dE/dx = w . s (MeV/ cm) (2-1)

Fig. (2-1): Dependence of the specific ionization s of alpha particles

on the depth x in the stopping material.

- One can conclude that while a parallel beam of mono-energetic α

particles are penetrating a matter their energy is decreased gradually

while their number remains constant up to the end of the track, where

they are converted into inert helium gas.

2-3 Interaction of beta particles with matter:

- Beta particles, which are electrons or positrons emitted in beta

negative or positive decay of some radio-nuclides, are lighter than alpha

particles by a factor of about 7360 times. So, the speed of beta particles

is higher than that of alpha particles with the same energy by a factor of

about 86 times. So, the speed of a beta particle with 1 MeV energy is close

to the speed of light (which is 3x108 m/s). These high speed of beta

particles together with their small mass lead to that they may loose a

considerable part of their energy not only through ionization and

excitation but also by completely different mechanism, due to the very

high de-acceleration of these particles near the atomic nuclei of the

s

R

2

7

matter. This mechanism is the emission of electromagnetic radiation (x-

ray) known as bremstrahlung radiation.

- As the velocities of beta particles are very high comparing with

alpha particles with the same energies, the interaction time between the

incident beta particle and the orbital electrons and the nuclei of the

atoms is very small, in comparison with the interaction time of an alpha

particle. Moreover, the beta particle and orbital electrons are of the same

mass. So, all these factors strongly affect the character of interaction

between beta particles and matter. The main discrepancies between beta

and alpha interaction with matter can be summarized in the following:

- Beta particles transfer their energy to the matter via two

mechanisms which are: ionization and excitation, and emission of

bremstrahlung radiation. At comparatively low energy of particles

(few hundreds KeV) the main process for energy loss is the ionization

and excitation. As the energy of these particles increases the contribution

of emission of bremstrahlung radiation increasesd ant at very high

energies, this contribution becomes the predominant process of energy

loss. Moreover, the role of emission of bremstrahlung radiation is

strongly dependent on the atomic number Z of the matter, where it

increases with the increase of Z. For this reason high Z material should

not be used for shielding sources. The best material that can be used to

shield sources are the light solid material, such as plastic or aluminum

to reduce the emission of bremstrahlung radiation (x-ray).

- The energy percentage f of beta particles, which is lost via the

emission of bremstrahlung radiation as a function of both beta particles

maximum energy Emax and the atomic number Z is determined as:

f = 0.035 Emax Z %

- The track of any beta particle in the matter takes the form of a

broken line (due to the similar mass of the two interacting particles).

- The energy transferred from the incident beta particle to the

orbital electron in a single collision varies from a very low portion of the

particle energy up be very high portion of this energy, so that the

complete energy of the incident particle may be transferred in a single

collision. This means that the delta electrons are predominant in

interaction with matter.

2

8

Fig. (2-2): The broken track of particles in the material

- The specific ionization s in beta interaction is much less than that

for alpha interaction (by a factor of about one hundred due to the smaller

interaction time). So the range of beta particles is much larger than that

of alpha particles. The range of 1 MeV particles is about 4- 5 m in air, 6-

8 mm in water, plastic or human tissue, and about 2- 3 mm in aluminum.

- Both particles (e.g. the electron and the positron) behaves in

the matter in accordance with the previously mentioned two

mechanisms, although they have different sign of the charge. However,

there is an essential difference between the two particles at the end of the

track. When the energy of the positron becomes very low, it annihilates

with one of the electrons of the matter, where they completely vanishes

as a mass, and these two masses are converted into electromagnetic

energy in the form of two photons, each with energy of 511 KeV. This last

process is known as the annihilation process and the two photons with

511 KeV are called annihilation photons.

- It is important to conclude that while a parallel beam of β

particles are penetrating a matter, not only their energies are decreased

as a function of depth in the matter, but also their number will be

decreased, due to two facts which are: (a) the continuous energy

spectrum of β particles, so that low energy particles will loose their

energy through, relatively, a very thin layer of the matter while high

energy particles can penetrate to much higher depth, (b) a large number

of β particles will be deflected from their initial direction due the their

broken track.

2

9

- Due to the above mentioned factors, the number of β particles

which penetrate a certain thickness of matter x is decreased

exponentially, in accordance with the following (2-2) relation:

N = N0 e – μ x

(2-2)

where N is the number of particles penetrating the thickness x,

N0 is the number of particles reaching the same point in the absence of

the absorber, and μ is known as the attenuation factor. This factor is

strongly dependent on both atomic number Z of the absorber and energy

E of the particles.

2-4 Interaction of x-ray and gamma radiation with matter:

- When a beam of x-ray or mono-energetic gamma radiation fall

on a matter, its photons may interact with this matter via one of the

following mechanisms, depending on the photon energy as well as on

the atomic number of the matter:

a- The photo-electric effect,

b- Compton scattering, and

c- The pair production.

- Other types of interaction between incident photons and the

matter, such as the interaction with the atomic nuclei, is considered

negligible from the point of view of radiation protection.

2-4-1 The photo-electric effect:

- In this process, the incident photon interacts with one of the

strongly bound orbital electrons of the atom (e.g. with any of electrons

belonging mainly to K or L shells, which are the closest shells to the

nucleus). In this type of interaction the photon delivers its total energy

Eγ to the orbital electron and completely vanishes, and correspondingly,

the electron will be knocked out from the atom, carrying an amount of

energy Ee equal to:

Ee = Eγ – B (2-3)

where, B is the binding energy of the electron in the corresponding shell,

defined as the amount of energy that should be delivered to the electron

just enough to liberate it from this shell (it varies from less than 1 to

about 100 KeV depending on the atomic number Z of the matter). If Eγ <

B, then the process will not occur. Correspondingly, the photo-electric

3

0

effect will yield one electron which carries approximately the photon

energy.

- The cross- section σph (sigma) of the photo-electric effect, which

is defined as the probability of occurrence of this effect, when a single

photon is incident on a unit area (1 cm2) containing a single atom,

strongly depends on the photon energy Eγ as well as on the atomic

number of the matter Z. This probability σph decreases very fast with

increasing the photon energy Eγ, while it increases very rapidly with

increasing Z, as Z4 up to Z5. The unit of σph is barn(1 barn = 10

-24 cm

2).

- Dependence of the photo-electric cross section σph on photon

energy Eγ is shown 0n figure (2-3) where the photon energy is expressed

in a logarithmic scale.

K-edge

σph

ln Eγ

Fig: (2-3): Dependence of the photo-electric cross section on photon

energy

2-4-2 Compton scattering:

- In this process, the incident photon interacts with one of the

very loosely bound orbital electrons of the atom, or with a free

electron (e.g. with any of electrons belonging to the outermost shells,

which are far away from the nucleus). In this type of interaction the

photon delivers a part of its energy Eγ to the electron and the photon

well be deviated (scattered) from its original direction, carrying the

remaining amount of energy. Correspondingly, the Compton scattering

3

1

of a photon will yield a photon with lower energy and a free Compton

electron, that carries the remaining amount of energy.

σc

ln Eγ

Fig: (2-4): Dependence of the Compton cross section on photon energy

- the cross-section σc of Compton scattering decreases

approximately slowly with increasing of the photon energy, while it

depends linearly on Z of the matter.

2-4-3 The pair production:

- In this process, the incident photon interacts with the strong

electric field of the atomic nucleus, when approaching it very closely

(e.g. interaction between the incident photon and the atomic nucleus),

and if the photon energy is higher than 1022 KeV. In this type of

interaction the photon vanishes completely, and one electron-positron

pair with rest mass equivalent to 1022 KeV is produced. If the energy of

the incident photon Eγ is higher than 1022 KeV, then the excess energy is

delivered to the produced electron and positron, in approximately equal

portions. Correspondingly, the pair production will yield two particles

which are the electron and the positron.

- The electron and the positrons behave inside the stopping matter

in the same way as beta particles, e.g. they loose there energy on

ionization and excitation of the atoms of this matter as will as on

emission of bremstrahlung radiation, depending on the atomic number of

the atoms of the absorbing matter. When its energy becomes very low

each positron annihilates with one of the orbital electrons, (e.g. this

positron and electron vanish as a mass converting into two photons, each

3

2

with energy of 511 KeV). These two photons may interact with matter via

photo-electric process or Compton scattering, or they both may escape

out from the matter without interaction, in a process known as a double

escape, or one photon may interact while the other may escape in a

process known as a single escape.

- The cross-section σp of the pair production process increases with

the photon energy increase. This increase is relatively slow after the

threshold value of 1022 KeV and becomes fast with increasing the energy.

This probability σp depends on the atomic number of the matter as Z2.

σp

1022 KeV ln E γ

Fig: (2-5): Dependence of the pair production cross section on photon

energy

- Due to the formation of energetic electrons and positrons,

resulting from the three processes of interaction between gamma

radiation or x-rays and the matter this radiation, is known as indirectly

ionizing radiation.

2-4-4 The total gamma cross section σ:

- The total gamma cross-section σ is defined as the total

probability for a single incident photon to interact with one atom

existing in a target of 1 cm2 when it collide this area via any of the three

processes, e.g:

σ = σph + σc + σp

3

3

- The unit of the total cross section σ is the barn (1 barn = 10-24

cm2).

2-4-5 The linear attenuation coefficient μ:

- By definition, the linear attenuation coefficient μ for a certain

matter and at a certain photon energy, is defined as the probability of the

interaction of a single photon that have this energy with all atoms

existing in a cube of 1 cm3 (1 cm

2 area and 1 cm depth) of this matter, on

which it falls by all the three processes. So, if the number of atoms in 1

cm3 is n, and the total interaction cross-section is σ, then it is clear that:

μ = n σ

σ

1022 KeV ln Eγ

Fig: (2-6): Dependence of the total cross section on photon energy

- The unit of the linear attenuation coefficient μ is cm-1

(e.g. per

cm). It is also clear from the behavior of σ as a function of the energy

that μ depends strongly on the atomic number Z of the attenuating

material, specially for both low and high energy photons. Moreover, μ is

strongly dependent on the photon energy Eγ.

2-4-6 The mass attenuation coefficient μm:

- In different references another physical quantity, known as the

mass attenuation coefficient μm is used instead of the linear attenuation

3

4

coefficient μ. This new quantity μm is defined by dividing the linear

attenuation coefficient μ by the density ρ of the attenuator, e.g:

m = μ / ρ

- It is seen that the unit of the mass attenuation coefficient μm is

(cm2/ gm). The reason for using μm instead of μ is that its value may be

considered, approximately, constant for different attenuating materials,

for the same photon energy.

2-4-7 The exponential attenuation of x and gamma radiation:

When a narrow beam of mono-energetic x-ray or gamma

radiation falls on a matter of thickness x cm, a part of the incident

number of photons No from this beam will interact with the matter via

any of the three known processes, resulting in the reduction of this

incident number as a function of the thickness x of the matter. Number

of the photons N, that will penetrate the thickness x without any

interaction with the matter will proceed in the same direction and do not

loose any part of their energies. This is expressed, mathematically, by

the following exponential law:

N = No e - μ x

- The exponential attenuation (e.g. exponential reduction of the

number of photons) is valid when specific conditions are applied. These

conditions are:

a) A very narrow beam consisting of parallel mono-

energetic photons.

b) A very small thickness x of the attenuator, so that,

multiple Compton scattering is negligible.

- In all other cases this exponential law is not valid due to

Compton scattering of photons from the broad beam as well as the

multiple Compton scattering of some photons due to the thick layer of

the attenuator. This will be discussed, in details, in a later chapter on

build-up.

- If the linear attenuation coefficient μ is used (in cm-1

) then the

thickness x of the attenuator should be expressed in (cm), to get non-

dimensional value of the product μ x. However, when the mass

attenuation coefficient μm is used (in cm2/gm), then the thickness of the

attenuator should be expressed in the so called mass-thickness xm, which

3

5

is obtained as the product of the linear thickness x of the attenuator and

its density ρ, e.g:

xm = x ρ

The unit of the mass-thickness xm is (gram/cm2).

- The exponential attenuation of x-rays and gamma radiation

makes the concept of the range for this type of electromagnetic radiation

is not valid. A definite portion of the incident beam will penetrate

through the attenuating matter, even when its thickness is too large. For

example, if a Co-60 source is shielded (surrounded) by more than 2 m

thick concrete wall some emitted photons from this cobalt will penetrate

through this shield, without suffering any kind of interaction.

2-4-8 The half value layer (HVL):

- The half value layer (HVL), or half value thickness, of a matter

at a certain gamma energy, is defined as the thickness of that matter,

which is necessary to attenuate the original number of the incident

photons No, with this energy, to its half value ( e.g. to N = 1/2 No). The

HVL is related with the linear attenuation coefficient μ with the

following simple relation:

HVL = 0.693 / μ

- Since μ is dependent on the radiation energy E and the material

of the attenuator Z, the HVL is also dependent on these factors.

- The unit of the HVL is cm when the μ is expressed in cm-1

, and

its unit is (gm/ cm2), when μ is expressed in cm

2/ gm.

2-4-9 The tenth value layer (TVL):

- The Tenth value layer (TVL), or Tenth value thickness, of a

matter at a certain gamma energy, is defined as the thickness of that

matter, which is necessary to attenuate the original number of the

incident photons No, with this energy, to one tenth of this value ( e.g. to

N = 1/10 No). The TVL has the same units as the HVL, and it is related

with last value with the following relation:

TVL = 3.32 HVL

3

6

2-4-9 The energy absorption coefficient μa:

- The energy absorption coefficient represents the portion of

energy absorbed from x-ray or gamma radiation in a definite volume of

the matter. This coefficient is used to account for the so called "kerma"

or absorbed dose from x or gamma radiation into the interacting matter,

(e.g. in dose calculations). It should be mentioned that authors of some

references are using, by fault, this coefficient to express the attenuation

coefficient μ. These Two coefficient (μa and μ, both linear and mass)

have different values, specially at medium and high photon energies, and

should not replace each other, except at very low photon energies (less

than few hundreds of KeV) where they are very close to each other.

- The reason of the discrepancy between μa and μ is the Compton

scattering and the pair production. In Compton scattering the photon is

deviated from its original direction, transferring only undefined part of

its energy to the matter, and the scattered photon may escape out from

this matter, so that although it has been omitted out from the beam, it

does not transfer its complete energy to the matter. In the pair production

the energy may not be transferred completely to the matter, since one or

even the two photons, resulting from the annihilation of the positron

with one electron may escape out of the matter.

- Due to the above mentioned reasons μ is almost higher than μa ,

specially with increasing the photon energy

2-5 Interaction of the neutrons with the matter:

- Since the neutrons are neutral particles (e.g. uncharged particles),

they do not interact neither with any of the orbital electrons nor electro-

statically with the atomic nuclei. They may interact only with nuclei via

nuclear forces, when they very closely approach any of them. This is the

reason of the high penetrating power of neutrons in the matter.

- the most important and efficient mean for energy transfer from

neutrons to the matter is the elastic scattering of the neutron on light

nuclei, such as hydrogen (in wax, water, polyethylene, or plastic),

deuterium (in heavy water) beryllium, carbon, and oxygen. With

decreasing the mass number of the interacting nucleus, the average

energy, transferred from the neutron to this nucleus, in a single collision,

increases. For this reason the hydrogen nuclei are considered the best

moderator for neutrons, and the materials which contain high

3

7

concentration of hydrogen, such as wax, water, Polyethylene, and plastic

are extensively used for effective slowing down of the fast neutrons. In a

single collision with a hydrogen nucleus, the neutron loses, in average,

63 % of its energy. This portion of energy is transferred to a proton,

which is the hydrogen nucleus.

- Since the recoil protons are heavy charged particles, they ionize

the matter. So, the neutrons are considered as indirectly ionizing

particles.

2-5-1 The neutron moderation:

- The neutron moderation means the slowing down of fast

neutrons (e.g. decreasing their energies from the MeV range to about

0.025 eV. Neutrons with such low energies are called thermal neutrons,

since their motion is controlled by the prevailing temperature.

- For slowing down of the fast neutrons (with energy of about

several MeV) to thermal neutrons, these neutrons should be subjected, in

average, to about 18-19 collisions with hydrogen nuclei. This number of

collisions requires a thickness of a hydrogen rich material, such as wax

or water of about 15- 25 cm.

- The thickness of the wax or water may be increased over the

mentioned values for radiation protection purposes, since these materials

absorb thermal neutrons with a certain probability forming deuterium

atoms which are stable.

- The role of inelastic scattering of neutrons for neutron

moderation is negligible.

2-5-2 The neutron capture:

- when a neutron approach very closely to a nucleus it may be

captured in it, forming a new isotope of the same element, with the

emission of a prompt gamma photon. An example of the neutron capture

reaction is:

no

1 + Cd114

47 Cd115

47 + γ

- The probability of the neutron capture is strongly dependent on

the neutron energy. The reaction cross-section (which represents the

probability of the neutron capture) increases strongly with the decrease

of the energy, reaching very high values for thermal and slow neutrons

3

8

(the slow neutrons are those with energies just higher than that of

thermal neutrons). Moreover, at certain energy values for the slow and

thermal neutrons, and for some nuclides the probability of the neutron

capture reaches very high values, known as a resonance neutron capture

or absorption. The energy values at which the resonance neutron capture

occurs depend on the absorbing nuclide. For example for Cd114

47 , it has

been found that the resonance capture occurs at thermal and low

energies, and the capture probability at resonance reaches extremely

high values. For this reason Cd114

47 is considered one of the best absorber

for thermal and slow neutrons.

- One of the most effective method to shield a neutron source and

to reduce effective doses around it is to put three layers of different

materials in the following consequence from the source: a) About 20 cm

of wax, plastic or any other solid (or liquid) material, rich with hydrogen

content to moderate fast neutron and convert them into thermal or slow

neutrons, then b) A thin sheet of Cd114

47 (with about 1 mm thickness) to

absorb thermal and slow neutrons, and finally c) a certain thickness of

lead to attenuate the prompt gamma radiation emitted in the neutron

capture in Cd114

47 .

- There are other materials that can be used practically to reduce

the neutron doses arising from different neutron sources, by moderation

and absorption of these neutrons, such as water (normal or light water),

boron and others

- In the absence of all of the mentioned materials one can use other

commonly existing materials in the field, such as the sand and other

types of soil. Although their shielding properties is too limited in

comparison with other materials, a large thickness of these sand or soil

may reduce neutron doses to lesser values due to the presence of some

light elements such as oxygen and carbon.

3

9

CHAPTER 3

RADIATION DETECTORS, SURVEY METERS

AND CONTAMINATION MONITORS

3-1 General:

- The main two processes which are used for detection of different

types of ionizing radiation are based on the use of:

a) Ionization of the detector material and formation of

electron-ion pairs, or electron hole pairs, and collection of this

charges or their current.

b) Excitation of the detector material and then measurement

of the emitted light during the de-excitation process, and

collection of this light or their current.

- There are other processes, which are used for detection and

counting of ionizing radiation. For example, one of these processes is the

use of activation of a certain nuclides by irradiation of certain material

by neutrons and then by measurement of the induced activity due to the

neutron capture.

- The type of the detector that should be used for detection and

counting and identifying of ionizing radiation depends strongly on:

a) The type of the radiation (e.g. heavy or light charged particles,

neutrons, x, or gamma radiation.

b) The energy of the measured particles or photons.

c) The intensity of the radiation field (e.g. the particle or photon

flounce).

d) The purpose of detection and measurement.

3-2 The gas detectors:

- In all gas detectors, detection of directly and indirectly ionizing

radiation is done through the ionization of some mixture of a gas

contained in a vessel with certain shape and volume.

- For directly ionizing radiation, such as heavy charged particles or

beta particles, the ionization of the gas atoms or molecules occurs inside

4

0

the detector vessel. The average number of the resulting primary

electron-ion pairs in the detector is defined by dividing the particle

energy (in eV) by 34 eV, which is the average energy needed to form

one electron- ion pair. For detection of heavy charged particles (such as

alpha), the detector wall should be equipped with a very thin window of

low Z material (less than 40 gm/cm2 of a light material) to permit the

entrance of these particles inside the detector, without loosing a

considerable part of its energy in this window. For the detection of beta

particles the window can be done from a thicker material, since the

range of these particles is much higher than that of alpha particles.

- For the indirectly ionizing radiation, namely x and gamma

radiation, ionization of the detector’s gas is done by the primary charged

electrons and positrons, emitted as a result of the interaction of the

incident photons with a very thin layer of a heavy material, such as lead,

fixed inside the wall of the detector. For detection of x and gamma

photons, There is no need to make a window in the detector wall due to

the very large range of photons.

- For neutrons, which are indirectly ionizing radiation too, the

ionization is done by charged particles such as protons emitted as a

result of the elastic scattering of the incident fast neutrons with hydrogen

nuclei existing in a very thin layer of polyethylene fixed inside the

detector wall, or by alpha particles, which are emitted as a result of the

neutron capture of thermal neutrons in certain gas materials with high

reaction cross-section, which is filling the detector, such as BF3 gas

(Boron tri-Fluoride) or others. Due to the high penetrability of neutrons,

there is no need to make any window in neutron detectors.

- There are three types of gas detectors which are:

a) the ionization chamber,

b) the proportional counter, and

c) the Geiger- Muller (GM) counter.

- For all types of gas detectors, the intrinsic detection efficiency

is 100 % only for all heavy charged particles. For beta particles the

efficiency is slightly less than 100 %, due to their continuous energy

spectrum, so that a part of the low energy particles will be absorbed

inside the window thickness. The efficiency of all gas detectors for

measuring photons or neutrons is extremely low, and strongly dependent

on their energy. For example the intrinsic efficiency of these detectors

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for photons may vary from few percents (2-4 %) to very low values (less

by many orders of magnitude) with increasing the energy of photons.

Remark: the intrinsic efficiency of a detector, for a certain type of

indirectly ionizing radiation at a certain energy, is defined as the ratio of

the number of particles or photons with the mentioned energy detected

by the detector from a given source, in a certain time period to the total

number of these particles or photons, with the same energy, incident

from the source on the detector surface, during the same time period. To

get the efficiency in percent this ratio should be multiplied by 100. For

example, if the intrinsic detector efficiency for photons with 662 KeV

energy is 2.5 % then this detector will detect only 2.5 % of photons

incident on its sensitive surface with this energy.

3-2-1 The ionization chamber:

- It is a detection device (see fig. (3-1), which consists of::

a- Two electrodes (anode a and cathode c) connected to a

moderate potential difference V (about 50- 100 volts depending

on the chamber volume and pressure) to secure collection of the

majority of the electrons and ions, which are generated by the

ionizing radiation inside the chamber on the anode and the

cathode respectively.

b- A guard grid g between the anode and the cathode to

secure independency the collected current, or consequently

voltage of the output pulse signal, resulting due to the passage of

this current through a high Ohmic resistance R, on the track

position of the incident particle.

- The ionization chambers can be used in a current regime (e.g. to

measure the very small average electric current, resulting by ionization

by a large number of incident particles or photons, and the chamber is

then known as a current type ionization chamber. They, also, can be

used to measure consequence pulses resulting from individual ionization

events (particles or photons), and hence to determine the number and

energies of these particles or photons, and in this case the chamber is

known as a pulse type ionization chamber.

- Since the collected current in the ionization chamber is too low

(in the range of pico-Ampers), the ionization chamber should be

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connected with a direct current amplifier (or pulse height amplifier) with

a very high amplification gain (thousands or more).

a C

g c V R

Fig (3-1): A diagram of an ionization chamber

- Ionization chambers are characterized by certain characteristics.

Some of these characteristics are:

a) The multiplication gain of any chamber equals 1, which

means that there is no multiplication of the electric current

resulting by ionizing radiation.

b) Relatively, high energy resolution r, which means that it

can be used to differentiate between particles or photons with

relatively close energies. The energy resolution of the ionization

chambers r varies between about 2.5 and 7 %, depending on its

volume and on the gas pressure.

Remark: the energy resolution r is defined as the ratio of the

energy fluctuation E caused by the detection process, to the

energy E of the particle multiplied by 100 (to get it as a percent)

e.g:

r = (E/E)x100 %.

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c) Relatively, a constant energy response curve in a wide

range of energies, comparing with all other detectors, when the

chamber is used as a detector in dose or dose-rate survey meters.

A constant energy response means that the ratio of the

measured dose (or electric current) from ionizing radiation with a

given energy E to that at a reference one Er remains constant in a

wide range of energies when the radiation field is homogeneous.

This is a very important property of ionization chambers.

d) In some cases the wall of the chamber is made from a

material having a similar composition as air to correct for energy

absorption in different materials, for more accurate determination

of doses or dose rates. In these cases the chamber is known as

air-wall ionization chamber.

e) For measurement of relatively high energy beta particle

or photons, it is necessary to increase the gas pressure inside the

chamber to secure full stopping of the ionizing beta particles

within it. In This case the chamber is known as a pressurized

ionization chamber. Such cambers are important for dose

measurements in a radiation field with a wide energy range.

- The shape of the output pulse from a pulse type ionization

chamber, which represents the detection of a single particle or

photon with a given energy value is demonstrated in fig.(3-2).

The polarity of th pulse on this figure is inverted, since it is

originally negative. The vertical axis shows the output voltage

amplitude of the pulse which is proportional to the energy of the

particle or photon, while the horizontal axis shows the time

duration of the pulse and dependence of its amplitude on time.

The voltage amplitude of the output pulses lies in the range of

less than one microvolt up to about one hundred microvolts,

depending on the particle energy. The pulse durations lies

between less than a 100 microseconds up to more than 1000

microseconds depending on the geometrical dimensions of the

chamber as well as on its internal capacitance and resistance. The

values of the used electronic devises such as the input impedance

and capacitance of the of this circuit strongly affect the duration

of the output pulses

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The pulse amplitude

The time (microsecond)

Fig (3-2): The pulse shape at the output of an pulse type

ionization chamber

3-2-2 The proportional counter:

- The proportional counter, (see fig 3-3) is a gas detector of a

cylindrical form, where a metallic cylinder is acting as the detector

cathode, while a very thin coaxial metallic wire with a regular diameter

is used as the anode.

- The applied voltage difference between the anode and the

cathode for the proportional counter is much higher than that used in an

ionization chamber with the same dimensions. This increase in the

applied voltage difference leads to the acceleration of ions and electrons,

so that they become capable to ionize new atoms, while they are moving

to the cathode and anode respectively. This yields in a high increase of

the electric current caused by ionizing radiations. So, the proportional

counter is acting as a detector and a current multiplier.

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V

Fig. (3-3): A diagram of a proportional counter

- The multiplication gain of the gas in the proportional counter

varies between about 100 to more than one thousand, depending on the

magnitude of the applied potential difference between its anode and

cathode.

- As a result of the multiplication the energy resolution r of the

proportional counter is much poorer than that of the ionization chamber.

Its values vary from about 10 to 30 %.

- Although the energy resolution of the proportional counters is

relatively poor, there is still some proportionality between the energy of

the detected particle or photon and the obtained current or pulse height

from this detector. This makes the accuracy of this detector for dose

measurements acceptable and this detector comes, directly, in the next

category after the ionization chamber, concerning the accuracy point of

view, as well as from the constancy of the energy response at relatively

wide range of photons energy.

- in spite of the relatively high multiplication gain in the

proportional counter, it still needs to be connected at the output to a

current or voltage amplifier, but with a lower amplification gain than

that used with the ionization chambers.

3-2-3 The Geiger- Muller (GM) counter:

- From the construction point of view the GM counters are exactly

similar to the proportional counters. The main difference is that the GM

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counter is operated at relatively higher potential difference between the

anode and the cathode.

- With increasing the applied voltage the current multiplication in

the gas of the tube becomes very high and almost reaches infinity. When

an ionizing particle or photon inters the GM tube, and when it interacts

with the detector material causing even one electron– ion pair a series of

consequent ionization occurs making avalanche multiplication. This will

cause occurring of electric discharge of the detector gas.

- The gas discharge will continue unless, it will be stopped by

internal or external reason in a process called quenching. The external

quenching is secured by inserting a large Ohmic resistance R in series

with the high voltage source, while the internal quenching is secured by

the addition of a certain ratio of a mono-atomic gas. The second

technique of quenching is preferred, since the first one leads to a serious

increase in the detector dead time, due to the increase of the magnitude

of the resistance.

- As a result of infinite amplification of the GM tubes, particles or

photons with different energies will give the same electronic signals with

the same pulse amplitude, so that, it can be measured without further

amplification.

- Due to the complete discharge through the detector tube, the

proportionality between the energy of the particle and the pulse height of

corresponding signal is completely lost. In other words the GM counter,

completely, does not differentiate between different energies, and it can

be only used to count the number of pulses (detected particles or

photons) independent of their energies.

- The dead time of a pulse type detector is defined as the time

period through which the electrons and ions are collected and treated as

a pulse. During the dead time the detector will not detect any other

ionization event, so If the time separation between two sequent ionizing

events (e.g. two consequent registered particles or photons) is less than

the detector dead time, then they will be detected as a single particle or

photon, and hence there will be some loss of the detected number of

particles or photons.

- The energy response curve of the GM counter is, comparatively,

worse than that of the proportional counter. For this reason, special

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filters are used with the GM counters to correct for the non-constancy of

the response curve.

- It should be mentioned that dose survey meters that use GM

counters as a detector, should not be used in any place containing high

radio-frequency (rf) source, such as linear accelerators, since they are

very sensitive to high frequencies and they almost give full scale reading

in these fields without the presence of any type of the ionizing radiation.

3-3 The scintillation detectors:

- In all scintillation detectors, detection of directly and indirectly

ionizing radiation is done through the excitation of some atoms, which

are consisted in a solid crystalline or liquid scintillator. So, any

scintillation detector, (see fig 3-4), consists, mainly, of, at least, two

components, which are:

- The scintillation crystal or liquid (the scintillator)

- The Photo-Multiplier Tube (PMT).

Fig. (3-4: The components of a scintillation detector

- Sometimes, there is a third component, which is the so called

light pipe. This pipe is made of a highly transparent type of silicon glass,

which is acting as a light conductor to transfer light photons emitted

from the crystal (or liquid scintillator) to the photo-cathode of the PMT.

The PMT

The light pipe The scintillator

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- All the components are matched together, without any air voids

or bubbles by putting a small drop of silicon oil between any of these

components and pressing so that no air bubbles are existing in between.

The detector components are enclosed inside a hermetically sealed

metallic enclosure, so that no light can penetrate through it.

- The function of the scintillator is to emit photons of visible light,

The number of these photons is linearly dependent on the energy of the

incident particle. As these emitted photons fall on the photo-cathode of

the PMT, a limited number of electrons will be emitted from this photo-

cathode. The number of these photo-electrons is linearly dependent on

the number of the incident photons on the photo-cathode, and

consequently, on the energy of the incident particle on the scintillator.

- The role of the photo-multiplier tube (PMT) is to multiply the

number of emitted electrons from the photo-cathode, by a very large

factor (at least some thousands times and much more). For this purpose

the PMT contains a large number of dynodes (about 9- 13 dynodes),

each of which is covered with a material with high coefficient of the

secondary emission. The emitted photo-electrons are accelerated toward

the first dynode by a positive voltage difference V, so that they gain an

amount of kinetic energy equal V electron volts, and become capable to

induce secondary electron emission from the dynode, so that their

number will be multiplied by a factor equal to the coefficient of

secondary emission . This coefficient is strongly dependent on the

voltage difference V and may reach, relatively, high values (up to 3 and

more) with the increase of V. Electrons emitted from the first dynode

are, again, accelerated toward the second dynode by another positive

voltage difference V, giving rise to another step of a secondary emission

from this second dynode, and yielding second multiplication . Then the

consequent acceleration processes toward the next dynodes with a

multiplication factors of on each one of these dynodes will yield a total

multiplication factor of n (if the value of is the same for all dynodes),

where n is the number of dynodes in the PMT. After multiplication a

huge number of electrons are emitted from the last dynode and they are

collected on the anode of the PMT, giving a negative pulse on the output

of this anode due to the presence of a high ohmic resistance.

- The anode pulse represents the registration of a single particle in

the detector, and the amplitude of this pulse is proportional to the energy

of the particle. So, the number of the registered pulses is proportional to

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the number of the incident particles or photons, while the amplitude of

each pulse represents the energy of the registered particle or photon.

Output pulses on the anode of the PMT have a similar form of the pulses

from an ionization chamber shown on fig. (3-2), but the time duration of

the pulse may be more less than that of the ionization chamber for some

types of scintillation crystals.

- It should be mentioned that the electron multiplication gain M of

the PMT, (which is approximately equal to the coefficient raised to the

power n (i.e. M n))

is strongly dependent on the biasing voltage V

which is supplied to the PMT Anode or cathode. This voltage is divided

by a potential divider using a set of resistances to bias the cathode, all

dynodes and the anode with the nominal voltages. It is recommended to

supply the PMT with the nominal voltage, since the increase of V will

increase the factor M, but at the same time it will shorten, strongly, the

service life-time of the PMT.

- different types of radiations are detected using different

scintillators. Table (3-1) represents the most widely used scintillators for

different types of radiations. All these scintillators emit violet light with

wave length shown in table (3-1).

- Alpha particles and protons can be easily detected using a thin

layer (about 1mm thickness) zinc sulphide crystal doped with silver ZnS

(Ag), while electrons and positrons can be detected using organic

crystals or liquids.

- The Sodium Iodide crystal with Thallium NaI(Tl) is the best

scintillation crystal that can be used to detect gamma radiation with a

higher efficiency, due to its high density. Moreover, the addition of a

small ratio of Thallium to the Sodium iodide makes the crystal capable

for emission of light photons at room temperature. To meat the required

detection efficiency of gamma radiation, the NaI(Tl) crystal is grown

with a different thicknesses. These crystals are available in the market,

mainly, in a cylindrical form with dimensions ranging from 1/2 inch

diameter x 1/2 inch height, up to more than 10 "

x 10 ". Generally

speaking, the scintillation gamma detectors are much sensitive to detect

gamma radiation, in comparison with gas detectors, and the detector

with 3" x 3

" NaI(Tl) crystal is considered as a reference one, so that, the

relative efficiency of any other gas and solid detectors, is given referring

to this reference one.

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- Fast neutrons can be easily detected by scintillation detectors

using secondary charged particles, which arise as a result of the neutron

elastic scattering or nuclear reaction. For example, these neutrons can be

detected by putting a very thin layer of polyethylene in front of the

ZnS(Ag) crystal, so that neutrons will collide with hydrogen atoms of

the polyethylene, yielding recoil protons, which are detected in this

crystal.

Table (3-1): scintillators used for detection of different radiations

Name and characteristics of the scintillation material Type of

radiation Name Physical form Density

(g/cm3)

Decay

time

(sec)

Wave

length

(nanometer)

Zink

sulphide

ZnS(Ag)

Solid crystal,

Low

transparency

3.67 1x10-5

450 Alpha and

protons

Sodium

iodide

NaI(Tl)

Solid crystal,

High

transparency

4.10 2.5x10-7

410 Gamma

rays

Anthracene Organic

compound

1.25 2.7x10-8

440 Beta

particles

Stylbene Organic

compound

1.15 4x10-9

410 Beta

particles

- Thermal neutrons may be detected either through using a lithium

iodide doped with thallium LiI(Tl) crystal as a scintillator, which has

characteristics close to those of NaI(Tl), or by using a mixture of lithium

or boron compound with the ZnS(Ag) crystal. Thermal neutrons interact

with the lithium or boron atoms of the crystal, giving rise to charged

particles, which, in their turn, cause the scintillation in the ZnS(Ag)

crystal.

- The energy resolution r, of different scintillation detectors

depends, mainly, on the volume of the used crystal, and with a lower

degree, on the characteristics of the used PMT. Small crystals have

better resolution r, while large ones are characterized with bad

resolution. The value of r varies between about 2.5 and 10 %, depending

on the volume of the crystal.

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- However the efficiency of the scintillation detectors for x and

gamma radiations is much higher than that of all gas detectors, its

response curve to dose variation with radiation energy is very poor,

comparing with all other detectors. For this reason scintillation detectors

are not widely used in different survey meters, for dose or dose-rates

measurements or.for radiation dosimetry, but they are very widely

used to search for a lost gamma source as will as for radiation

counting and spectroscopic measurements as well as in surveying

ground resources of nuclear ores.

3-4 The Semi-conductor detectors:

- The semi-conductor materials used in manufacturing electronic

devices and radiation detectors are the silicon and germanium. Both of

these elements have tetravalent atoms, and their crystalline structure is

formed, so that, each atom has a covalent bond with four neighbor

atoms. When radiation interacts with one of these atoms an electron of

the four outermost electrons is ejected, and it becomes free, and then its

atom is left without an electron, This free of electron place is known as a

hole. So, while interacting with a silicon or germanium crystal radiation

will generate electron-hole pairs. The energy required to form one

electron-hole pair in silicon is about 1.1 eV , in average, while the

energy required for germanium is about 0.7 eV. For this reason, the

number of electron-hole pairs formed in silicon by a particle or photon

with certain energy is higher than the number of electron-ion pairs

formed in an air ionization chamber by a factor of about 30 times for

silicon and of about 48 times for germanium. As a result of that, the

energy resolution of semiconductor detectors is much better than that of

the ionization chamber. For example, the energy resolution r for a

germanium gamma detector with a cylindrical crystal of about 60 mm

diameter and 60 mm height is about 1.75 KeV for the 1332 KeV gamma

ray line of Co-60, (which is about 0.13 %).

- At present, hyper-pure germanium crystals of different shapes

and volumes are produced for use as a powerful tool for high resolution

gamma ray spectrometry in the fields as well as in the fixed laboratories,

to meat the required efficiencies. Their relative efficiency cover a very

wide range starting from about 10 % up to more than 100 % with respect

to the 3" x 3" NaI(Tl) detector. The only disadvantage of these detectors

is that they required a very deep cooling, prior to their operation, and

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this is achieved, mainly, through cooling by liquid nitrogen (- 189 oC) or

by electric cooling.

- Silicon surface-barrier detectors are produced, since the sixties of

the 20th

century up to now, with different shapes and thicknesses, to

detect heavy charged particles of different energies, and they are widely

used in high resolution alpha spectrometry, as well as in spectroscopic

measurements of heavy charged particles (such as protons, deuterons

and others). Their energy resolution is as good as about 0.4- 0.5 %, and

they do not require any cooling.

- Other pure silicon crystals are produced to be used for high

energy resolution spectroscopic measurements of x-ray and low energy

gamma radiation up to about 100 KeV. These detectors, again. require

the deep cooling as germanium ones prior to their operation.

3-5 The survey meters:

- Surveying the radiation areas and measurements of radiation

doses and dose rates is one of the required activities that must be

conducted, regularly, in all areas, where radiation sources are used, and

around these areas, to evaluate the radiation levels and, consequently to

assess the radiation doses to the occupational workers as well as to the

general public.

- There is no single survey meter, which can be used to survey all

types of radiations, and the choice of the survey meter is strongly

dependent on the type of radiations or particles, their energy, as well as

on their intensity.

- Any survey meter consists, mainly, of:

a- A radiation probe or detector, which is assigned for a

certain type of radiation, and for a certain range of energy, as

well as for a certain range of radiation intensities ore doserates,

b- An electronic circuit for current or voltage amplification.

c- A measuring device to measure the amplified electric

current or to count the pulse rate or the number of pulses during a

defined time interval.

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d- A devise, which convert the current intensity or the pulse

rate or the number of pulses to dose rate or accumulated dose

through the defined time.

e- Some types of survey meters are equipped with a sound

device that gives sound clicks as an larm indicating pulse

counting. This is essential to demonstrate by sound the radiation

level, without the need to look to the scale of the survey meter.

- Any survey meter should be characterized by a constant relative

response curve over the whole energy range existing in the surveyed

radiation field. The relative energy response of a survey meter is defined

as the ratio of the current intensity at different energies, to that current

intensity at a certain definite energy (or the ratio of the pulse number per

unit time at different energies to the pluses number per unit time at a

certain definite energy), when the radiation field is homogenous and

constant. Fig. (3-5) represents the relative response curves for an

ionization chamber (curve a), GM counter (curve b), and NaI(Tl)

scintillation detector. From this curve it is easily seen that the ionization

chamber is characterized by a relatively constant response curve, in the

energy range from about 100 KeV, up to about 2 MeV, while the GM

counter, and specially the scintillation detector, have a strongly varying

response with energy. With respect to the GM counter, better response

may be attained by using a set of filters, made from different materials

such as lead and others.

3-5-1 Calibration of the survey meters:

- Survey meters used for determination of dose ore dose rates

arising from beta particles, gamma radiation and x-rays and neutrons

should be recalibrated periodically, each six months, depending on the

prevailing working conditions. For example, in practices of industrial

radiography, which may lead to serious radiological hazards, it is

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Fig. (3-5): Energy response for some detectors

An ionization chamber.

A GM counter.

A scintillation detector.

required to carry out the recalibration each six months, while for other

practices with lower source activities, the recalibration may be repeated

yearly.

- The recalibration should be conducted, only, by recognized and

authorized laboratories, and by qualified persons from the national

regulatory authority. The recalibration should cover all ranges and scales

of the survey meter. Moreover, each scale should be recalibrated, at

least, at two points, to assure the accuracy in the full range of the scale.

A recalibration certificate should be issued, showing the date of

recalibration, the name of the specialist, who conducted it, the

recommended date for the next recalibration, and comments about the

constancy of the calibration constants of the device.

- The recommended radiation sources for calibration of different

devices are:

a) X-ray machines with proper high voltages for calibration

of survey meters used with x-ray sources.

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b) Cesium-137 and Cobalt-60 sources for calibration of

gamma survey meters.

c) Sr-90 sources for calibration of beta survey meters.

d) Am-Be source or Cf-252 sources for calibration of

neutron survey meters

3-6 The contamination monitors:

- A contamination monitor is a device used to detect

contamination of surfaces, hands and feet, clothes and surface

contamination on equipments with any radio-nuclides. There are other

contamination monitors that are used to detect contamination of air with

radio-nuclides, such as iodine monitors, which are, widely, used in

laboratories of the nuclear medicine in the hospitals. Any contamination

monitor should be able to detect very small contamination (up to 185 Bq

or 0.005 micro-Curie). For detection of a lower contamination, another

procedure, known as wipe test, should be conducted.

- To detect surface contamination with any radio-nuclide (except

Tc-99m), only alpha or beta particles should be detected, since gamma

radiation and neutrons have a very high penetration power, and hence,

they will be detected independent of their location inside the sealed

container or on the external surface of this container. For this reason, any

contamination monitor consists, mainly, of:

a- An alpha or beta particle detector, which is prepared with

a very thin window to permit these particles to pass to the

detector to be detected inside it.

b- An electronic amplifier circuit for voltage amplification,

to get measurable pulses.

c- A measuring device to count the pulse rate or the number

of pulses during a defined time interval.

d- The contamination monitors are, always, equipped with a

sound device that gives clicks indicating pulse counting. This is

essential to demonstrate by sound the contamination level,

without the need to look to the scale of the monitor.

- To detect surface contaminations with alpha emitters or with beta

emitters, with relatively high beta particles energy, using a wipe test, an

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appropriate gross alpha beta counter should be used for high

contamination levels.

- For detecting surface contaminations, of low contamination

levels, with alpha emitters or with low energy beta emitters, such as

tritium (H-3), using a wipe test, a liquid scintillation detector should be

used for increasing the solid angle and the detection probability.

- The wipe test, for detection of any contamination on the outside

surface of any radio-active source should be conducted, periodically,

each six months. If the counting facilities needed to detect the surface

contamination of the sources are not available at the licensee, then he

should contract a qualified and recognized party to conduct these tests

on behave of him.

3-7 Devices for personal dosimetry:

- In all controlled areas monitoring of the personal doses of the

workers must be done using, internationally, recognized personal

devices, such as either the Thermo-Luminescent Dosimeter (TLD) or the

Film badge. In the Kingdom of Saudi Arabia., the TLD are the

recognized device.

- Some of the widely used TLDs are the lithium fluoride (LiF) or

calcium fluoride (CaF) non-metal crystals. When the ionizing radiations

fall on any of these crystals and interact with orbital electrons of their

atoms, some of the electrons are transferred from the, so called,

equivalence band to a higher band. One of the main characteristics of

these crystals is that the transferred electrons remain in the new band at

the prevailing temperature. When the crystal is heated up to a

temperature of 200 Celsius, the transferred electrons return back to their

original band, with the emission of a visible light. The amount of the

emitted light linearly depends on the amount of radiation energy

absorbed in the crystal. So, measurement of the amount of the emitted

light using a device, such as a photo-multiplier tube, is a good indication

of the amount of energy delivered from the ionizing radiation to the air

or the human body.

- The CaF crystal is characterized by a high sensitivity to

radiation, however its energy response is limited, while the LiF is

characterized by a good energy response, but its sensitivity is limited.

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- Another, internationally recognized, personal dosimeter is the, so

called "film badge", consisting a plastic film covered with a silver

bromide (AgBr) emulsion. When ionizing radiations interact with the

emulsion some electrons are ejected out, breaking the covalent bond

between silver and brome. When the film is processed the silver atoms

are collected in dots forming some darkness in the plastic film. This

darkness is a measure of the amount of radiation energy, to which the

film was exposed. To differentiate between different radiations and

energies, different filters are used between the film and its badge.

- The advantages of the film badge, in comparison with the TLD,

is that it is much cheaper, and it is considered as a document, since the

darkness remains for long time, together with its simplicity. Its

disadvantages, with respect to the TLD, is its limited accuracy and the

need to isolate it from the direct light.

_ In the supervised areas personal doses may be evaluated by

measuring the maximum dose rates in the place where the workers are

working, and by registering the total time during which the worker is

existing inside the area. In this case the dose rate should be measured as

the maximum value between the head and the knee.

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CHAPTER FOUR

SOME RADIOATION MEASURMENT TEQNIQUES

AND STATISTICAL FLUCTUATIONS

4-1 Introduction:

- In this chapter a very limited number of radiation measurement

techniques, including both relative and absolute measurements, together

with some factors affecting the accuracy of these measurements, will be

touched.

- Our discussion will be limited to the case when the radiation

source is located outside the radiation detector. In this case, there are

different source-detector configurations, described as good or bad

geometrical configurations, depending on the source and detector sizes,

as well as on the distance R between the source and the detector.

- The good geometrical configuration is defined as that one, at

which the size of the source can be considered as a point, and the source

detector distance R is than the largest dimension of the detector, so

that the different rays emerged from the source toward the detector are

approximately parallel in the detector. For realistic situations, with some

approximation, the good geometrical configuration is considered that

one, in which the source-detector distance R is, at least, ten times larger

than the larger dimension of the source or the detector. For radioactive

sources, with relatively low activity, it is impossible to realize a good

geometrical configuration, since the number of particles or photons

emitted from the source toward the detector will be too limited, so that

the measurement can consume a very long time, or it cannot be carried

out due to the existence of an intensive background radiation, which may

much exceed the intended radiation emitted from the source. In these

conditions, the measurements can be conducted, with a good accuracy,

in the so called bad geometrical configuration. So, one can conclude that

the source detector configuration is determined by many factors, such as

size and shape of the source and the detector, the source activity, the

type of radiation, and purpose of measurements. Fig. 4-1 represents

some of the experimentally used source-detector configurations both

good and bad.

- The advantages of a good geometrical configuration is that the

detector solid angle , (fig. 4-1 a) through which the detector sees the

5

9

source, can be easily and accurately calculated. Moreover, in this

configuration the detector intrinsic efficiency is constant and does not

change with relatively small variations of the source location.

4-2 The solid angle :

- Consider an isotropic point source with activity A Bq at a certain

distance R from the detector (fig 4-1 a). Since the particles (or photons)

are emitted from the source with equal probability in all directions, only

a small portion of these particles (or photons) have a chance to fall on

the detector surface. This portion is equal to the ratio between the

detector sensitive surface area, which faces the source, and the area of

the sphere, on which the detector surface is located and center of which

is the source, and the radius of which is R. In other words the solid angle

is defined as:

number of particles emitted per second inside the space

defined by the contours of the source and detector aperture

=

number of particles emitted per second from the source in

all directions

- So, the solid angle for a point isotropic source and a detector

with a circular aperture with radius r, located at a distance R from the

source, in a good geometrical configuration is given, in general, as:

= r2

/ 4R2

- Suppose you have a Cesium-137 source with activity 1

microCurie, located at a distance of 40 cm from a detector with a

circular sensitive cross-section with radius 2 cm. Then the detector solid

angle is:

= (2)2 / 4 x (40)

2 = 6.25 x 10

-4

- By multiplying the source activity A in becquerels by the

element of the solid angle , the number of particles, which reach the

detector each second is defined. For example, when the previously

mentioned source is used with a detector in the mentioned configuration,

from the 37000 particles emitted from the source in all directions, in

each second, only 37000 x 6.25 x 10-4

= 23 particles will fall on the

sensitive detector surface (provided that no particle will be absorbed in

the air between the source and the detector).

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0

The Source R The detector

(a) Good geometrical configuration

The source

The source The detector

The detector

(b) (c)

The source inside

a marinelli beaker

The detector A well type detector

(d) (e)

Configuration (b,c, d, and e ) are belonging to a bad geometrical configuration

Fig. (4-1)

6

1

- If the activity of the alpha or beta source is A Bq then it emits A

alpha or beta particles per second in all directions. If the resultant

daughter radionuclide is a gamma emitter, and if the probability of

emission of a gamma ray line with a definite energy E is f, then the

number of photons emitted with this energy in all direction, in one

second, is defined as (f A) photons. For example, it is known, that f =

0.85, for the 662 KeV gamma ray line emitted from barium 137, formed

by the beta decay of Cs-137. Then the number of 662 KeV photons

emitted from 1 Ci Cs-137 source is = f A = 31450 photons per second.

From this last number only 31450 x 6.25 x 10-4

= 19.66 photons will

impinge on the sensitive area of the detector in the configuration

mentioned previously.

- Finally, one can defined the total number N of photons with

known probability f, that impinge on the detector surface from a source

with activity A, when the detector solid angle is as:

N = f A

4-3 The detector intrinsic efficiency :

- In case of the indirectly ionizing radiation, such as gamma

photons and neutrons, the detector can only detect a certain portion of

photons or particles, incident on its sensitive surface. The detector

intrinsic efficiency is defined as:

number of photons detected in the detector per second

=

number of photons impinging on the detector per second

- For photon detectors, there are different intrinsic efficiencies,

such as the photo-peak intrinsic efficiency ph, Compton intrinsic

efficiency c, and full peak intrinsic efficiency f.

- All intrinsic efficiencies are, strongly, dependent on the energy

of detected photons or particles, the type and dimension of the detector,

the density of the material used as a detector, and in some conditions on

the nature of the source, its density and dimension. So, the efficiency

determined for a certain detector and at a given radiation energy must

not be used for another detector or at other radiation energy or for a

source of other density or shape.

6

2

4-4 Relation between the counting rate C and source activity A:

- The counting rate C of a counting system is the number of counts

(photons or particles) detected by this system in one second 9In the SI of

units). So, using the definition of the intrinsic efficiency , and the

number of photons or particles N = f A, impinging on the detector

sensitive surface, it is then clear that

= C / f A

- So, the source activity A is related with the counting rate C,

efficiency of the detector, the solid angle and f value of the certain

gamma line with a simple relation, which is:

C = f A

- In a bad geometrical configuration this relation is not used, due

to the large uncertainties in calculation of even by using very

complicated computer programs, as well as in determination of the

detector intrinsic efficiency . For this reason, another quantity known as

the overall efficiency is introduced, which combines both intrinsic

efficiency and element of the solid angle together, i.e: = . The

relation between the counting rate C, the overall efficiency and the

source activity then is:

C = f A

- When this last relation is used the overall efficiency is

determined experimentally, using a standard source with an accurately

determined activity A. the obtained efficiency is used to determined

activities A of unknown sources, provided that all sources are measured

in the same geometrical configuration (i.e sample volume, shape,

location and density) and using the same energy lines.

4-5 Other factors affecting the measurements:

- In performing relative and absolute measurements there are many

other factors, which may, strongly, affect the accuracy of the obtained

results. Some of these factors are the source itself, the medium between

the source and the detector, and the detector itself.

6

3

4-5-1 Role of the source effects:

- One of the important effects of the source is the self absorption

inside the source. The size, and in particular the way the source is made

may have strong effect on the measurement results. Whether the source

is a solid material or a thin deposited evaporated layer on a metal disc

this may make a difference. For gamma and neutron measurements, the

effect of the source thickness is relatively limited, while it is very strong

in measurement of charged particles, and specially alpha particles. In all

cases, self absorption factor fa in the source should be taken into

consideration, since it reduces the experimentally determined activity.

For this reason, sources of the charged particles (especially alpha)

should be very thin (not more than few micrograms/cm2

- The second important effect of the source is the backscattering

effect on source backing. The source is, always, deposited on a metallic

backing or support. This backing may lead to the scattering of particles,

(especially beta particles). The particles which are directed from the

source toward the backing may suffer backscattering, so that they will be

reflected back to the detector, increasing the count rate over the real

value. The backscattering factor fb, strongly depends, on the atomic

number Z of the backing material, backing thickness X and kinetic

energy E of the particles. Increasing Z, X or E will strongly increase the

backscattering factor fb. For this reason the source backing should be

made from a material with low Z and should have the minimum

thickness. The backscattering may increase the real count rate by a value

up to 70 – 80 %.

4-5-2 The role of the medium:

- The effect of the medium between the source and the detector is

important, too, in some circumstances. Normally the medium between

the source and the detector is air, which have a very low density. For this

reason this medium will have ignored effect on the measurement results

for photons and neutrons. If the measured particles are charged, then all

the particles suffer some energy loss, and some of them (especially those

with low energy) may be completely absorbed, while some others may

be scattered in or out of the detector. To eliminate these effects, source

of the charged particles and detector should be placed in an evacuated

chamber.

6

4

- To demonstrate the role of the medium, it is important to remind

that alpha particles with energy 5 MeV loose completely, their energy

during about 4 cm of air, while beta particles with an end point of about

1 MeV loose their energy in the air within a layer of 4- 5 meters.

4-5-3 The role of the detector:

- In most cases the source is located outside the detector. The

radiation must penetrate the wall of the detector in order to be counted.

Interaction between the impinging radiation on the detector window and

the material from which this window is made, may scatter and (or)

absorb some of the impinging charged particles and even low energy

photons. This will lead to lowering the measured count rate with respect

to the real one.

4-6 Dead- time correction:

- the dead time , or resolving time of a detector, is defined as the

minimum time that can elapse between the arrival of two successive

particles at the detector, so that two distinct pulses are produced. The

most important is the dead time of the system as a whole which may be

composed of a detector, preamplifier, amplifier, ADC (Analogue-to Digital

Converter) and the multi-channel analyzer (MCA). However, since the

dead time of the detector is much longer than that of the electronics, the

later may be ignored (except the dead time of the MCA). So, the total or

detector dead time , as appropriate, should be taken into consideration

when counting ionizing radiation..

- As a result of the dead time, some pulses are not produced in the

detector, or not registered in the MCA, so that they are lost. The effect of

lost counts will be particularly important in the case of high counting

rates. Obviously, the measured counting rate should be corrected for the

loss of some counts due to the dead time. When the counting rate is too

high, then the system will stop functioning (counting) and it seems to be

dead all the time.

- If the dead time of the system (or the detector) is seconds, and

the experimentally measured counting rate is C (count per second), then

the fraction of time during which the system was dead is C second.

When the product C is 1, then the system will stop counting. The

6

5

relation between the true counting rate Ctr, the measured count rate C,

and the system dead time is:

Ctr = C / (1 – C )

- For clarifying the role of the dead time suppose that the dead

time of a system is 400 microseconds (s), and that the measured

counting rate is 30000 counts per minute (cpm). In this case the system

will be dead for 400 x 10-6

x ( 30000 / 60) = 0.2 seconds during the one

second, which means that the percentage of the dead time is 0.2 x 100 =

20 %. The true counting rate is then:

Ctr = 30000 / (1-0.2) = 30000 / 0.8 = 37500 counts per minute.

- For more clarification, suppose that the same system will be used

to register a measured counting rate of 150000 (cpm). In this case the

system, during one second, will be dead for 400 x 10-6

x ( 150000 / 60) =

1 second, which means that the percentage of the dead time is 1 x 100 =

100 %, i. e. the system will be dead all the time and it will stop counting.

4-7 The statistical fluctuation of radiation measurements:

- the radioactive decay is a truly random process, since no one can

predict when a certain single atom will be subjected to decay. So, all

radiation measurements are subjected to two types of errors which are

statistical and systematic errors. The next paragraphs will deal briefly

with some concepts of the statistical errors, which should be described in

statistical terms. The random processes obey the, so called, Gaussian

distribution.

- According to this Gaussian distribution, the standard error in a

mean value n of a set of readings ni, consisting of N readings, is defined

in terms of the, so called, standard deviation of the distribution. This

deviation is defined as:

= [(1/ N) ( ni – n )2 ]

1/2

- In practical situations, scientists are making a single

measurement rather than many measurements to determine the true

mean. In this case if the number of the detected events is m then the

standard deviation of this value may be given as:

= (m)1/2

6

6

So, any measured value is reported as m = m (m)1/2

.

- the percentage statistical error E is defined as:

E % = ( / m) x 100 = 100 / (m)1/2

%

So, it is seen that the percentage statistical error E % decreases

when the amount of the measured counts m is increased. This fact is

represented in table (4-2), showing the number of the registered counts

in each measurement, together with its standard deviation and

percentage error in three cases known as; lower, medium and higher

confidence level, corresponding to 1 , 2 , and 3 respectively.

Table (4-2)

Number of

counts in the

reading

The standard

deviation

The percentage Error %

1 2 3

1 1 100 200 300

4 2 50 100 150

16 4 25 50 75

25 5 20 40 60

100 10 10 20 30

400 20 5 10 15

1000 31.6 3.16 6.32 9.48

10000 100 1 2 3

100000 316 0.316 0.632 0.948

1000000 1000 0.1 0.2 0.3

- To get acquainted with the so called lower, medium and higher

confidence level, suppose that a certain experiment with a long lived

radioactive isotope, such as uranium (with a half life time of 4.468 x 109

years), have been repeated 1000 times, with a mean count of 400. the

long half life is intended to be sure that no change has been occurred

during the 1000 measurements. In this case according to the laws of the

statistical distribution, the counts measured in these thousand runs will

be as shown in table (4-3)

4-7-1 The standard error in the counting rate:

- In practice, the number of counts recorded in the presence of a

given source, is usually recorded either in a scaler or as the total (gross)

counts G in a peak of interest in a multi-channel analyzer, during a

6

7

certain time period of measurement tg. However, the reported result is

the counting rate, i.e., counts recorded per unit time (namely per

second), which is Cg = (G / tg). In some cases, especially, when dealing

with a gamma source, the amount of the background gamma radiation

may be comparable with that radiation emitted from this source, and it

should be taken into consideration, to get the net count rate Cnet from the

source.

For this reason, the background counts B should be measured, in

the absence of the source, during an appropriate time period tb, and the

background count rate Cb = (B / tb) is determined. To get the net count

rate Cnet, resulting from the source alone, the background count rate Cb

should be subtracted from the gross count rate Cg, i.e.,

Cnet = Cg – Cb = ( G / tg ) – ( B / tb )

Table: (4-3)

The range of the experimental readings among

the thousand readings

The number of readings

380 – 420 which meat (m ) 680

360 – 380 which lie between [(m-2) and (m-) 136

420 – 440 which lie between [(m+) and

(m+2)

136

340 – 360 which lie between [(m-3) and (m-

2)

23

440 – 460 which lie between [(m+2) and

(m+3)

23

Less than 340 1

More than 460 1

The lower confidence level includes all reading higher than (m-1) i.e higher

than 380 or lower than (m+1), i.e. lower than 420. These are 680 +136 + 23

+1 = 840 readings among the thousand, with 84 % confidence

The medium confidence level includes all reading higher than (m-2) i.e higher

than 360 or lower than (m+2) i.e. lower than 440. These are 680 +136 +136 +

23 +1 = 976 readings among the thousand, with 97.6 % confidence.

The higher confidence level includes all reading higher than (m-3) i.e higher

than 340 or lower than (m+3), i.e. lower than 460. These are 680 +136 + 136

+ 23 + 23 +1 = 999 readings among the thousand, with 99.9 % confidence

6

8

- The standard deviation net in the net count rate Cnet is defined as:

net = [G /( tg)2 + B /( tb)

2 ]

1/2

- To reduce the error which may arise due to the background

radiation in measurements of low activity gamma sources, both the

source and the detector are placed inside a special shield.

6

9

CHAPTER FIVE

DOSIMETRY QUANTITIES AND THEIR UNITS

- The quantities used to measure the dosimetrical quantities of

ionizing radiation are based on the gross number of this radiation in a

defined situation or, on the gross amount of energy, deposited in a

defined mass of material.

5-1 The Exposure:

- The exposure is defined as the exposure of the dry air at standard

temperature and pressure (STP), to x-rays or low energy gamma ray (up

to 3 MeV).

- The old unit of the exposure is the ROENTGEN – R

- The SI (Systeme Internationale) unit of the exposure is Coulomb

per Kg dry air.

- The Roentgen is defined as the amount of exposure of the dry air

at the standard temperature and pressure, which yields a charge of 2.58 x

10-4

Cooulomb/Kg dry air of each sign (Electrons or ions).

5-2 The absorbed dose D:

- The fundamental dosimetric quantity in radiation protection is

the absorbed dose. The absorbed dose D is defined as the ratio of the

amount of energy E deposited from the ionizing radiation to a mass

element of m of a matter e.g:

D = E / m

In other words, the absorbed dose is defined as the amount of

energy deposited from any type of ionizing radiation, into a unit mass of

any matter.

- the units of the absorbed dose are the "rad" in the old (CGS

which is Centimeter Gram Second) system of units, and the "Gray Gy"

in the standard international system of units.

- The rad is the old unit of the absorbed dose (e.g. in CGS system

of units). One rad is defined as 100 erg of energy deposited into one

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0

gram of matter. The word rad is the abbreviation of a sentence which is

"radiation absorbed dose”..

1 rad = 100 erg / 1 gram

- The Gray "Gy" is the SI unit of the absorbed dose (e.g. in MKS

system of units, Meter, Kilogram, Second). One Gray is defined as 1

Joule of energy imparted into one Kilogram of matter.

- It should be mentioned that the absorbed dose is defined in terms

that allow it to be specified at a point, but it is used to mean the average

dose over a tissue or organ.

- Using the relation between Joule and erg which is 1 Joule = 107

erg, it is clear that

1 Gy = 100 rad

5-3 The equivalence between the Roentgen R, rad and Gy:

1 R is equivalent to 87 erg/gm = 0.87 rad = 0.0087 Gy in air

1 R is equivalent to 96 erg/gm = 0.96 rad = 0.0096 Gy in the human

tissue

- So, one can consider, with acceptable approximation that:

1 R 1 rad 0.01 Gy

5-4 The Kerma K:

- The Kerma K is defined as:

K = Etr /m

Where, Etr is the sum of the initial kinetic energies of all

charged ionizing particles, liberated by the uncharged ionizing particles

in a material of mass m.

- The units of the Kerma are the same units of the absorbed dose.

- It should be mentioned that the Kerma is approximately equal

the absorbed dose at very low photon energies, but it becomes less than

the absorbed dose, at medium and relatively high photon energies.

5-5 The Radiation Weighting Factor WR:

7

1

- It has been found that the probability of the so called stochastic

effects depends not only on magnitude of the absorbed dose, but also on

the type and energy of radiation delivering this dose. This is taken into

account by weighting the absorbed dose by a factor related to the quality

of the radiation for causing health effects. In the past, this weighting

factor has been applied to the absorbed dose at a point and was called the

quality factor Q. The weighted absorbed dose by the Q factor was called

the dose equivalent.

- The radiation weighting factor WR is a multiplier of the

absorbed dose to account for the relative effectiveness of different types

of radiation in inducing health effects. The values of this factor for

different types and energies of radiation are given in table (5-1).

Table (5-1): The values of WR

Type and energy range of radiation Radiation weighting factor

WR

Photons, all energies 1

Electrons and muons, all energies 1

Neutrons, energy < 10 KeV

10 KeV to 100 KeV

> 100 KeV to 2 MeV

> 2 Mev to 20 MeV

> 20 MeV

5

10

20

10

5

Protons, other than recoil protons, energy >

2 MeV

5

Alpha particles, fission fragments and

heavy nuclei

20

5-6 The Equivalent dose HT:

- It is the absorbed dose averaged over a tissue or organ T, due to

radiation of type R, and weighted by the radiation weighting factor WR ,

e.g:

HT = WR * DT, R

- When the radiation field is composed of different radiation types

with different values of WR , the equivalent dose to this tissue is given

as:

7

2

HT = R

WR * DT, R

- One should differentiate between the equivalent dose HT in a

tissue or organ and the dose equivalent H, which was used by the ICRP

before 1990. The dose equivalent H represented the dose in a point

rather than in a tissue or organ, since the quality factor Q, used for

weighting, represented this factor at the point.

- The units of the equivalent dose are the "rem" in the old CGS

system, or the Seivert "Sv" in the SI system.

- The "rem" (roentgen equivalent man) is the unit of the

equivalent dose in the old CGS system, where the absorbed dose is

measured in "rad".

- The Seivert "Sv" is the unit of the equivalent dose in the SI

system, where the absorbed dose is measured in Gray "Gy".

5-7 The tissue weighting factor WT:

- The relationship between the probability of stochastic effects and

equivalent dose is found, also, to depend on the tissue or organ

irradiated. The factor which represents the relative contribution of a

certain tissue or organ to the total detriment, from a uniform irradiation

of the whole body is called the tissue weighting factor WT.

- The tissue weighting factor WT is a multiplier of the equivalent

dose HT to an organ or tissue, to accounts for the different sensitivities of

different tissues and organs to the induction of stochastic effects. The

recommended by the ICRP values of the tissue weighting factors are

given in table (5-2)

5-8 The effective dose E:

- The effective dose E is defined as the sum of the weighted

equivalent doses in all the tissues and organs of the human body. In

other words it is defined as the sum of the tissue equivalent doses each

multiplied by the appropriate tissue weighting factor. It is given by the

Expression:

E = T

WT * HT

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3

- The units of the effective dose are the same units which are used

for effective dose in both CGS, and SI systems, i.e: the "rem" and the

Seivert "Sv' .

- Both effective and equivalent doses are quantities intended for

use in radiation protection including the assessment of the risks in

general terms. They provide a basis for estimating probability of

stochastic effects, for absorbed doses well below the thresholds for

deterministic effects.

Table (5-2): Tissue weighting factors averaged over the sexes and ages

Tissue or organ Tissue weighting factor WT

Gonades 0.20

Red bone marrow 0.12

Colon 0.12

Lung 0.12

Stomach 0.12

Bladder 0.05

Breast 0.05

Liver 0.05

Oesophagus 0.05

Thyroid 0.05

Skin 0.01

Bone surface 0.01

The remainder 0.05

The whole body 1.00

5-9 The committed equivalent or effective dose:

- Following an intake of a radio-nuclide into the human body,

there is a period during which this nuclide gives rise to equivalent doses

in the tissues of the body at varying rates. The time integral of the

equivalent dose rate is called the committed equivalent dose H().

Where is the integration time in years following the intake. If is not

specified, it is considered to be 50 years for adults and 70 years for

children.

- The committed effective dose E() is similarly defined as the

committed equivalent dose.

7

4

- Both committed equivalent dose and committed effective dose

have the same units as equivalent or effective doses.

5-10 The collective equivalent or collective effective dose:

- All the dosimetric quantities mentioned before relate to the

exposure of an individual. However, other quantities related to the

exposure of a group of workers or population are necessary. These

quantities are the collective equivalent dose ST, in a certain tissue of a

group of people and the collective effective dose S of this group.

- The collective equivalent dose ST is the equivalent dose

incurred in a defined tissue or organ by a group of workers or by a

critical group of people. The collective equivalent dose is defined as

the product of the number N of exposed individuals and their

average equivalent dose HT, when the amount of this dose is equal

for each member of the group.

- The collective effective dose S is the effective dose incurred by

a group of workers or by a critical group. The collective effective dose

is defined as the product of the number N of exposed individuals and

their average effective dose E.

- If several groups are involved, in the exposure, then the total

collective effective dose is the sum of the collective doses for all groups.

- The old unit of the collective equivalent dose or collective

effective dose is the "man.rem", while the SI unit is man.Seivert"

man.Sv".

- The collective quantities represent the total consequences of

exposure of a population or group of workers. For example, when the

risk factor R for lethal cancer is given, it is easy to assess the additional

cancer deaths, induced by radiation, among a group of exposed people

by multiplying their collective dose and the risk factor R.

7

5

CHAPTER SIX

BIOLOGICAL EFFECTS OF RADIATION

6-1 Direct and indirect actions of ionizing radiation on cells:

- The gross biological effects, resulting from exposure to ionizing

radiation are due to long and complex series of events, which are

initiated by ionization or excitation of relatively few molecules in the

organism. For example, the lethal dose for 50 % of the exposed men (or

women) within 30 days (LD-50/30) is known to be about 4 Gy of

gamma rays. This high and lethal dose affects only 1 atom from each 10

millions atoms.

- It is known that most of the human body is water, and most of

the direct action of radiation is therefore on water. The result of energy

absorption from radiation by water molecules is the production of

highly reactive free radicals, which are chemically toxic, and may exert

their toxicity on other molecules (a free radical is a fragment of a

compound or an element, that contains an unpaired electron). When a

water molecule is irradiated, it may become ionized, i.e;

H2O H2O+ + e

- (a physical stage which occurs within 10

-15

seconds from the moment of irradiation).

The positive ion dissociates immediately according to:

H2O+ H

+ + OH

and the electron will be captured by a neutral water molecule forming a

negatively charged water molecule according to the reaction:

e- + H2O H2O

-

This last negative ion dissociates immediately as:

H2O- H + OH

-

The last three reactions are known as the physio-chemical stage,

and they occur within 10-6

second of the moment of irradiation. The

positive and negative ions H+ and OH

- are of no consequence, since all

body fluids contain significant concentrations of them. The free radicals

H and OH may combine with like radicals, or with other molecules,

specially in case of irradiation of the human body with high LET

(Linear Energy Transfer) particles, such as alpha particles or fast

7

6

neutrons. This combination yield a hydrogen peroxide molecule, which

is a stable compound and capable to diffuse far from the point of its

generation.

OH + OH O2H2

Moreover H2O2 is a very powerful oxidizing agent and can affect

cells and molecules that did not suffer radiation damage directly. If the

irradiated water contains dissolved oxygen molecules, then the free

hydrogen radical may combine with the oxygen, to form the hydro-

peroxyl radical:

H + O2 HO2

This radical has a greater stability and can diffuse away and

combine with a free H radical to form hydrogen peroxide, thereby

further enhancing the toxicity of the ionizing radiation.

- Our knowledge is still too limited concerning the gross of

biological effects, that may occur long after irradiation.

6-2 Radiation effects:

- As a result of the processes discussed in the previous paragraph,

the living cell may be damaged. The most important damage is that

which may occur in the DNA. Damage in the DNA may prevent the

survival of the cell, affect its reproduction, or modify the cell itself.

- If enough cells in the organ or tissue are killed or prevented from

functioning normally, there will be a loss of the organ function, which is

known now as a deterministic effect. The loss of function will become

more serious as the number of affected cells is increased. Many organs

and tissues are not affected by small reductions in the number of the

available cells and the body will attempt to repair this damage.

However, if the decrease is large enough, then the body cannot repair

the damage, and the end result will be the death.

- The response of the body to develop a clone of modified somatic

cells is complex. The development of such a clone may be inhibited,

unless it is promoted by an additional agent, before or after irradiation,

and the clone may be eliminated or isolated by the body’s defenses.

However, if it is not, it may result after a prolonged or variable delay,

called the "latency period", in the development of a malignant

7

7

conditions in which the proliferation of modified cells is uncontrolled.

Such condition are grouped together and called cancer. A modified

germ cell in the gonads will transfer genetic information to the

descendants of an exposed individual, which may cause severe harm to

some of these descendants, known as hereditary effects. The somatic

and hereditary effects are known as "stochastic effects".

- There is some experimental evidence that radiation appears to

enhance immunological responses and to modify the balance of

hormones in the body, thus strengthening the natural defense

mechanisms of the body. Most of the data on such effects termed

"hormesis" have been inconclusive because of statistical difficulties at

low doses.

6-3 Deterministic and stochastic effects:

- The deterministic effects are the radiation effects for which a

threshold level of dose exists, and above which the severity of the effect

is greater for higher doses. Deterministic effects occur from acute doses

and some of these effects are radiation disease, cataract, erythema and

others. Deterministic effects are belonging to prompt somatic effect,

which means that they appear promptly after the threshold on the

exposed person.

- The stochastic effects are radiation effects, generally occurring

without a threshold level of dose (i.e they may occur from low doses as

well as from high doses), and their probability is proportional to the

incurred dose, and their severity is independent of the dose. Examples

of the stochastic effects are different cancers, leukemia‟s, and hereditary

effects. The stochastic effects are considered as delayed effects.

6-4 Acute deterministic effects:

- Acute exposure is defined as any single exposure to high dose of

radiation, during a short period of time, and which produce biological

effects within a short time after exposure so, they are called prompt

effects. All deterministic effects arise due to acute exposures. These

exposures may lead, also, to stochastic effects. The acute radiation

syndrome is subdivided into three classes. In the order of increasing

severity, these are:

7

8

a) The hemopoietic syndrome.

b) The gastrointestinal syndrome

c) The central nervous system syndrome

Certain diseases or effects are common to all these classes, which

are grouped under one name as "radiation disease or sickness", which

includes nausea and vomiting, malaise and fatigue, increased

temperature, and blood changes.

- The blood changes is reflected in the changes of the blood count.

These changes, usually, do not appear before gamma ray doses of 250-

500 mGy, but they, certainly, appear after 500 mGy. The white blood

cells known as leucocytes, which are counted in healthy adults as

7000/mm3 of blood are responsible for combating the infecting

organisms. There are two main types of the leucocytes, which are

granulocytes and lymphocytes, with relative proportion of 70- 75 %

and 30- 25 % respectively. The granulocytes are produced in the red

bone marrow and circulate for about 3 days before death, while

lymphocytes are produced in the lymph nodes and spleen, and remain

alive for 24 hours. After an acute exposure in the sub-lethal range there

is a sharp increase in the number of granulocytes, followed within a day

by a very sharp decrease to reach the minimum for several weeks or

months after exposure. The lymphocytes drop sharply after the

exposure, and remain depressed for several months.

- The hemopoietic syndrome appears after a gamma dose of about

2 Gy. This disease is characterized by depression or ablation of the red

bone marrow. The onset of the disease is, rather, sudden, and is heralded

by nausea and vomiting within several hours after overexposure. At 4-6

Gy complete ablation of the bone marrow occurs. An exposure of about

7-8 Gy or greater leads to irriversable ablation of the bone marrow. The

LD-50/30 is in the range of 3-5 Gy.

- The gastrointestinal disorders may appear from relatively small

doses (about 1-2 Gy) due to the death of a part of cells of the intestinal

epithelium, but the syndrome is severe after about 10 Gy. This

syndrome is a consequence of the desquamation of the intestinal

epithelium, and its signs are severe nausea, vomiting and diarrhea,

which begins very soon or immediately after exposure and the death

within 1-2 weeks is the most likely outcome.

7

9

- Central nervous system syndrome occurs, after relatively high

dose of acute exposure which is not less than 20 Gy. Its sign is the

occurring of unconsciousness, within minutes after exposure, and the

death occurs during several hours to few days.

- The skin may be subjected, due to its location to more radiation

exposure, especially in the case of low energy x-ray and beta particles.

An exposure of the skin to about 300 R in the diagnostic x-ray

(approximately 3 Gy) results in erythema, while higher doses may cause

pigmentation, blistering and ulceration.

- The gonads are particularly radiosensitive. A dose of about 150-

200 mGy to the tests in a single exposure results in temporary sterility

among men, but in case of prolonged exposure the dose rate threshold is

0.4 Gy/year The corresponding values for permanent sterility are about

3-6 Gy for acute exposure and 2 Gy/year for prolonged one. For women,

the threshold for permanent sterility is an acute absorbed dose to the

ovaries, in the rang of about 2.5- 6 Gy.

- The threshold for opacities of the eye lens (cataract), which occur

after some delay, seems to be in the range of about 5-10 Gy for an acute

exposure to low LET radiation. For high LET radiation the absorbed

dose threshold is 2 -3 times less.

6-5 The stochastic effects:

- As it has been mentioned, all cancers and hereditary effects

belong to the stochastic effects. For these effects there is no threshold

for their induction. As a somatic effect in humans, the period between

exposure to radiation and recognition of a cancer lasts a number of

years, known as latency period. The median latency period seems to be

about 7- 8 years for leukemia, while it seems to be two three times

longer for many solid tumors, such as breast or lung cancers. However

there are some types of cancers that may appear after about two years

latency.

- Up to now, Epidemiologic data on carcinogenicity of low doses

of radiation are contradictory and inconclusive. For this reason it is

prudent to estimate the risk probability at low doses by extrapolation

from the probabilities at high doses. At high doses, there are many

evidences that the cancer dose response is linear or linear-quadratic, for

human beings and for some other biological species. So, at present, the

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0

accepted evaluation model for excess cancer deaths due to radiation is

based on a linear zero threshold for the extrapolation to low doses. A

typical dose to cancer relationship for low and high-LET radiations is

illustrated on fig. (6-1).

Fig. (6-1): A typical dose effect relationship for low and high- LET

radiations

- The excess mortality from all cancers, attributed to a collective

effective dose of 1 man.Sv, in an acute uniform whole body exposure by

low LET irradiation, (or in other words the risk factor R, which

represents the probability of death from induced cancer by radiation per

1 man.Seivert) is illustrated in table (6-1). These values were obtained

and corrected by different national and international scientific

committees, as indicated in this table.

- The relative probabilities of fatal cancers in different organs and

the total Risk factor for Japanese population, sexes averaged, ages 0- 90

years, 0- 19 years, 20- 64 years, Evaluated by Japan and the National

Institute of Health (NIH) of USA are tabulated in table (6-2). It is

evident that the relative probabilities vary with age group by a factor of

about 2 for leukemia and colon cancer.

8

1

Table (6-1) The excess mortality from all cancers,

attributed to a collective effective dose of 1 man.Sv,

Source of estimate The risk factor per 1 man.Sv

Additive model Multiplicative model

BEIR I, 1972 1.2 x 10 -2

6.2 x 10 -2

UNSCEAR, 1977 2.5 x 10 -2

-

BEIR II, 1980 0.8 x 10 -2

– 2.5 x 10 -2

2.3 x 10 -2

- 5.0 x 10 -2

NUREG, 1985 2.9 x 10 -2

5.2 x 10 -2

UNSCEAR, 1988 4.0 x 10 -2

– 5.0 x 10 -2

7 x 10 -2

– 11 x 10 -2

BEIR V, 1990 - 8.85 x 10-2

6-6 Hereditary effects

- Two kinds of radiation induced genetic damage, when one of the

two parents is irradiated, are considered important. These two kinds are

gene mutations (alterations in the genes) and gross chromosome

aberrations (alteration in the structure or number of the chromosomes).

This damage may be transmitted and become manifest as hereditary

disorders in the descendants of the exposed individual.

- Hereditary effects vary widely in their severity. When the

production of dominant mutation occur, it may lead to genetic diseases,

predominantly in the first and second generation progeny after exposure,

and they may be seriously harmful and life- threading. Chromosomal

aberrations may also result in congenital abnormalities in children.

Moreover, Interaction of genetic and environmental factors may leads to

the so called multi- factoral disorders.

- For low doses and dose rates, the ICRP estimates the nominal

hereditary effect probability coefficient for severe effects (excluding

multi-factoral effects), related to the gonad doses, and over all

generation, to be about (0.6- 1.1) x 10-2

per man.Sv.

- The principal effects of irradiation on the mammalian fetus

include:

a) Lethal effects in the embryo.

b) Malformations and other constructural changes.

c) Mental retardation.

d) Introduction of malignancies including leukemia.

e) Hereditary effects.

8

2

Table (6-2): Relative probabilities of fatal cancers in different

organs and the total Risk factor for Japanese population

Organ

The relative probability

Multiplicative model NIH

0- 90 y 0- 19 y 20- 64

y

0- 90 y 0- 19 y 20- 64

y

Oesophagos 0.038 0.021 0.061 0.042 0.024 0.063

Stomach 0.291 0.266 0.305 0.268 0.225 0.301

Colon 0.180 0.255 0.089 0.121 0.171 0.066

Lung 0.174 0.191 0.159 0.221 0.297 0.129

Breast 0.023 0.025 0.022 0.027 0.034 0.019

Ovary 0.014 0.009 0.023 0.019 0.013 0.025

Bladder 0.052 0.030 0.082 0.052 0.028 0.080

Bone marrow 0.077 0.052 0.109 0.100 0.055 0.165

Remainder 0.150 0.050 0.150 0.150 0.150 0.150

All cancers 0.999 1.000 1.000 0.998 1.000 1.000

total probability

(10-2

per man.Sv)

10.7

24.6

7.8

9.7

21.5

7.3

- Lethal effects can be induced in experimental animals by small

doses (0.1 Gy) immediately after implantation of the embryo into the

uterine wall. Human pregnancy loss is known to occur following

exposure to ionizing radiation. However, data on the probability of fetal

death, for different dose are sparse, and it is difficult to conclude on the

dose- effect relationship coefficient.

- Malformation can arise spontaneously, as well as from radiation.

The most dangerous period of exposure is during the most active phase

of cell multiplication and differentiation in the structure of the concern.

However, malformation may also occur at all stages, especially in the

later phase of pregnancy. The malformations appear to result from the

killing of cells. Dose-effect relationship is not derived for humans, and

the corresponding relationship for animals may be applied.

- Mental retardation was not observed to be induced by radiation

prior to 8 weeks from conception, or after 25 weeks. During the most

sensitive period, 8-15 weeks after conception, the fraction of those

8

3

exposed which became severely mentally retarded increased by

approximately 0.4 per Sv. For exposure during weeks 16-25, it

increased by about 0.1 per Sv. Moreover mental impairment of lower

severity is also apparent in children exposed in uteri. Evidence of such

impairment is still being collected.

- Irradiated fetuses seem to be susceptible to childhood leukemias

and the childhood cancers, which are expressed during, approximately,

the first decade of life. The risk of fetal childhood cancer due to prenatal

exposure has been estimated to be 2.8 x 10 -2

Per Sv. Constancy of risk

throughout the pregnancy was assumed.

8

4

CHAPTER SEVEN

DOSE CALCULATION

7-1 Dose calculation from point sources:

- Effective dose rates from point exposed sources, existing outside

the human body, which emit beta particle, gamma radiation and

neutrons can be easily calculated with high accuracy, using very simple

mathematical relations. For alpha emitters the external irradiation of the

human body is, completely of no importance, since the energy of alpha

particles is completely absorbed in about 4 cm of air. Even, if a person

is very close to an alpha emitting source, the emitted particles will be

fully absorbed in a very thin layer of the dead skin. However, the

radionulides which emit alpha particles are considered the most

hazardous particles if these radionuclides are ingested or inhaled or

taken by any other mean of intake,(internal exposure), due to the high

linear energy transfer from these particles and, correspondingly, due to

the high specific ionization and high radiation weighting factor of heavy

charged particles.

7-2 Dose calculation for the beta emitters:

- For any beta emitting source, which is relatively small in

dimension, so that it may be considered as a point source, and which is

very thin enough, to neglect the self absorption inside the source, the

dose rate Ė (in microSeivert/hour Sv/h) resulting at a point p, existing

at a distance d (in meters) from the source center can be calculated with

a good accuracy using eq.(7-1):

Ė (Sv/h) = (5 A Eav) / d2 (7-1)

where Eav is the average energy of the beta particles emitted from

the beta source (in MeV), and A is the source activity in Mega

Becquerell, MBq). If the average energy Eav of the beta emitting

radionuclide is not known, then it can be easily approximated with

accepted degree of accuracy from the maximum energy Emax (i.e. end

point energy) of this radionuclide as; Eav = Emax /3 .

- When the relation (7-1) is used to calculate the dose rate Ė of

beta emitters, the distance d between the source and the point of interest

should be limited enough (up to about 2-3 m), to ignore the absorption

8

5

of beta particles in air, which may play a considerable role, specially

when the emitted particles are of relatively low energy.

- When the average energy, Eav of the beta particles is relatively

small, then energy absorption in the source material, air and detector

window should be taken into consideration. The percentage of the

absorbed energy in these media should be subtracted from the dose rate

calculated by equation (7-1).

7-3 Dose calculation from external gamma sources:

- For a relatively small gamma emitting source, so that it may be

considered as a point source, and which is relatively thin, to neglect the

self absorption inside the source, the dose rate Ė (in microSeivert/hour

Sv/h) resulting at a point p, which exists at a distance d (in meters)

from the exposed source, which emits a single gamma ray line (i.e. with

a single energy) can be calculated with a good accuracy using eq.(7-2).

Ė (Sv/h) = (0.142 A f E) / d2 (7-2)

where, A is the activity of the source (in Mega Becquerell MBq),

E is the gamma ray energy in (MeV), and f is a factor representing the

ratio between the number of gamma photons with the defined energy E

emitted per each 100 decays of the radionuclide..

- If the source emits more than one gamma ray line (i.e. it emits

gamma ray with different fixed energy values Ei , then the dose rate is

calculated using equation (7-3);

Ė (Sv/h) = 0.142 A ∑i (fi Ei) / d2 (7-3)

where, the summation ∑i is taken for all gamma ray lines i , and

the product fi Ei represents the product of fi for the i line, and its energy

Ei .

- It is faster to calculate the effective dose rate Ė from any exposed

gamma source if the gamma specific constant (or factor) for this

source is known. The gamma specific constant (or factor) for a certain

radionuclide (in the SI system of units) is defined as the dose rate (in

Sv/h), at a distance of 1 m from the source of this radionuclide,

activity of which A is one Mega Becquerell. When this constant is

available then equations (7-2) and (7-3), for a single line emitters or

multiple lines emitters will look as equation (7-4).

8

6

Ė (Sv/h) = A / d2 (7-4)

- Comparing equations (7-2) and (7-3) with (7-4), it is clear that

the gamma specific constant for a gamma emitter which emits a single

gamma ray line is:

= 0.142 f E (7-5)

While, the gamma specific constant, , for a radionuclide which

emits multiple gamma ray lines is:

= 0.142 ∑i (fi Ei) (7-6)

and the quantities are defined in the same manner as in equation

(7-2) and (7-3).

- It should be mentioned that the unit of the gamma specific

constant (or factor) , in the SI system of units is (Sv. m2/h. MBq). In

this system of units the gamma specific factor is defined as the dose

rate at adistance one meter of a source, nactivity of which is one Mega

Becquerel. In the classic system of units, the gamma specific constant

is defined as the dose rate (in Roetgen/hour R/h), at a distance of 1 m

from a source of a radionuclide, activity of which A is one Curie (Ci).

So, the unit of in the classic system of units is (R. m2/ h. Ci). Up to

now, some books and references are using the classic system of units.

For this reason, one should be capable of transferring this constant

between the two systems of units. For this purpose, equation (7-7)

represents the relation between them.

(Sv.m2/h.MBq) = (R.m

2/h.Curie) / 3.7 (7-7)

- Table (7-1) gives the values for the gamma specific factor for

some widely used radionuclides in some practices, in the two systems of

units.

7-4 Dose calculation from neutron sources:

- It should be mentioned that all neutron sources, used in different

practices, are emitting fast neutrons with a continuous spectrum (i.e.

with energy varying from some tens of KeV up to 7-9 MeV), and the

neutron yield is isotropically distributed. The neutron generators, which

are used in different applications emit monoenergetic neutrons. In the

absolute majority, these generators emit neutrons, as the result of the

(d,n)reaction on tritium, with 14.1 MeV energy.

8

7

- Neutron sources, with isotropic neutron distribution, may be

considered as a point source. The dose rate Ė (in microSeivert/hour

Sv/h) resulting at a point p, which exists at a distance d (in meters)

from the source, can be easily calculated using eq.(7-8).

Ė (Sv/h) = (0.08 C N) / d2 (7-8)

Table (7-1): the gamma specific factor for some radionuclides

The radionuclide The gamma specific constant

(Sv.m2/h.MBq) (R.m

2/h.Curie)

Cesium Cs137 0.087 0.325

Cobalt Co60 0.356 1.32

Gold Au198 0.0622 0.23

Iodine I-131 0.0595 0.22

Iridium Ir198 0.13 0.48

Radium Ra226 0.223 0.825

Sodium Na24 0.497 1.84

where, C is the neutron to effective dose rate conversion factor,

which is tabulated in some references, N is the number of neutrons

emitted from the source per second.

- Table (7-2) gives the values of the conversion factor C, for some

some neutron energies, in a unit to get the neutron effective dose rate in

(Sv/h).

Table (7-2): the gamma specific factor for some radionuclides

The neutron

energy

C The neutron

energy

C

1 KeV 3.74 x 10-6

1 MeV 1.32 x 10-4

10 KeV 3.56 x 10-6

5 MeV 1.56 x 10-4

100 KeV 2.17 x 10-5

10 MeV 1.47 x 10-4

500 KeV 9.25 x 10-5

8

8

7-5 The Inverse square law for external exposure:

- This law is applicable to all gamma and neutron point sources.

For beta particle sources, the law may be applied, only for relatively

small distances d, due to the absorption of a fraction of beta particles

energy in air. This law states that, the effective dose rate Ė from a point

source inversely depends on the square of the distance d between the

source and the intended point. This means that when the distance from

the source is doubled the dose rate will decrease four times, and when

the distance is increased three times, the dose rate will decrease 9 times

(32 = 9). In mathematical representation, when there are two points from

a source, located at distances d1 and d2 , and the dose rates at these

points are Ė1, and Ė2 respectively, then the inverse square law states

that, the dose rates and their distances are related by the following

equation:

Ė1 d12

= E2 d2 2 (7-9)

7-6 Dose calculation from internal exposure:

- When radionuclides are taken inside the human body via

ingestion, inhalation, or through the skin, the exposure, then, is called

internal exposure and the effective or equivalent dose arising due this are

described as committed doses. Any element or compound, when it is

ingested or inhaled inside the human body behaves in a different

manner, and its metabolic behavior depends on many factors including

dietary habits of the human beings. The simplest and most accurate

approach to evaluate the effective committed dose E, (in Seivert) from

the internal contamination with a certain radionuclide, via ingestion or

inhalation can be easily done using equation (7-10), in case of intake of a

different radionuclides I the total comotted dose is determined as::

E (Sv) = i Ci Ni (7-10)

where Ci is the dose conversion factor of the intake of one

Becquerell of a radionuclide i (in Seivert/Bq), and Ni is the activity of

the itaked amount of the radionuclide i (in Bq). The summation should

be taken for all these radionuclides. (by ingestion or by inhalation,

depending on the pathway),

8

9

- It should be mentioned that the dose conversion factor Ci

strongly depends on the pathway of the intake (ingestion, inhalation or

through skin), as well as on the chemical and physical form of the

intaked radionuclide and on the solubility of the chemical compound in

which it is existing. Moreover the factor is strongly age dependent. For

this reason The ICRP, IAEA and UNSCEAR have Published these

factors separatelely for ingestion and inhalation and for different age

groups (less than 1 year, from 1- 2 years, from 2- 7, from 7- 17, and

adults). The obtained committed doses using these conversion factors

refer to the dose incurred up to 70 years age.

7-7 The Annual Limit on Intake ALI:

- It is the intake by ingestion, inhalation or via the skin, of a given

radionuclide in a single year, by the reference man (70 kg mass), which

would result in a committed effective dose equal to the relevant annual

dose limit for the occupational workers. The ALI is expressed in the unit

of activity.

- For occupational workers, where the dose limit, now, is 20 mSv,

the ALI for a certain radionuclide i (in Bq) can be determined, using the

dose conversion coefficient Ci for this radionuclide, (in Sv/Bq) which

represents the committed effective dose per intake of 1 Bq, via the

defined mode of intake by the following (7-11) relation.

ALIi (Bq) = 20x10-3

/ Ci (7-11)

- It should be mentioned that intake of 1 ALI by any mean of

intake in one year is equivalent to an effective committed dose of 20

mSv/year

7-8 The Derived Air Concentration (DAC)

- It is defined as the maximum concentration (in Bq/m3) of a single

radionuclide I in the air at the working place of an occupational

reference man, breathing of which during the whole working hours

through the year (2000 hours/year), would result in a committed

effective dose equal to the relevant occupational dose limit (i.e 20 mSv).

- The DAC of a certain radionuclide I is derived using the volume

of the breathed air during the work hours. According to the used model,

9

0

the reference man breathes, in average, 1.25 m3 of air during one

working hour, if the work is of a moderate type. This means that the

inhaled volume of air during the 2000 working hours/year is 2500 m3.

So, the (DAC) is defined in terms of ALI, as:

(DAC)I (Bq/m3) = (ALI)I / 2500 (7-12)

- It should be mentioned that inhalation of 2000 DAC hours by

any worker in one year is equivalent to an effective committed dose of

20 mSv/year

9

1

CHAPTER EIGHT

RADDIATION SHIELDING

8-1 Shielding of sources of alpha particles:

- It has been mentioned, in chapter 2, that alpha particles, emitted

from all their natural sources, are absorbed within a relatively thin layer

of air (about 4 cm for 5 MeV particles). So all external sources of alpha

particles do not require any shield, provided that there are no other types

of radiation (beta, gamma or neutrons) emitted from them.

8-2 Shielding of sources of beta particles:

- To make a proper shield for a beta source one should use, only,

light rigid material, with low atomic mass number Z, since high Z

materials interact with these particles, yielding a considerable portion of

highly penetrating x rays. The portion f (in percent) of beta particle

energy transferred to emit bremstrahlung radiation (i.e x-ray) is defined

from the maximum energy of the beta spectrum Emax (in MeV), and the

atomic number Z of the interacting material as:

f = 0.035 Z Emax % (8-1)

So, to make a proper shield for a beta source one should use,

only, light rigid material, with low atomic mass number Z, since high Z

materials interact with these particles, yielding a considerable portion of

highly penetrating x- rays

- Although beta particles are characterized by a continuous energy

spectrum, their mass range Rm in any matter can be easily calculated

using the maximum energy Emax of the beta spectrum. For this purpose,

the mass range Rm in (gram/cm2) is defined as the linear range R (in cm)

in the defined material, multiplied by its density , i.e:

R m = R . (8-2)

- The mass range Rm of beta particles in a given material is defined

as a function of the maximum energy Emax of the beta spectrum from the

given radionuclide, using the following (8-3) relation, provided that Emax

is expressed in MeV:

Rm (gm/cm2) = 0.412 Emax (1.265 - 0.0954 ln Emax) (8-3)

9

2

- Equation (8-3) is applied when the beta particles maximum

energy is in the range between 0 up to 2.5 MeV. At higher beta particles

energies other equation is used for determination of Rm, which is:

Rm (gm/cm2) = 0.53 Emax - o.106 (8-4)

- It should be mentioned that the thickness Rm of the shield for

beta particles, which is sufficient to, completely, absorb these particles

does not depend on the activity of the source, so that the shield which is

sufficient for any small activity is also sufficient to shield any other

much larger activity of the same beta emitter. This principle is not

correct for x-ray or gamma radiation, where the source activity is the

most important factor in determination of the thickness of the shield.

- Due to the energy loss of some beta particles energy via

emission of the bremstrahlung radiation (x- ray), the shields of all beta

sources, which must be made from low Z material should be surrounded

by another shield, which is made of a high Z material to absorb the x-

rays, emitted during the interaction of the beta particles with the beta

shield. Calculation of the later shield will be covered in the next

paragraph.

8-3 Shielding of x and gamma ray sources

- In chapter 2, the linear attenuation coefficient μ for a certain

matter and at a certain photon energy as well as the mass attenuation

coefficient μa have been defined and used in an exponential form to

express the attenuation of the number of x-ray or gamma photons as a

function of the thickness x of the material. Number of the photons N,

that will penetrate the thickness x without any interaction with the matter

was expressed, mathematically, by the exponential law:

N = No e - μ x

- This exponential attenuation (e.g. exponential reduction of the

number of photons as a function of x) is valid for calculation of the

thickness of the shield for electromagnetic radiation, only, when the

beam of parallel mono-energetic photons is very narrow, and the

thickness x of the attenuator is very thin.

9

3

- In order to calculate the effective dose rate Ė of gamma radiation

due to a certain shield, one should use the mass-energy absorption

coefficient μa instead of the mass attenuation coefficient μm, due to the

reasons, mentioned in chapter 2. So, the relation between the dose rate Ė

in the presence of the shield of a thickness x and dose rate without this

shield Ė0 is:

Ė = Ė0 e – μa x

(8-5)

- In all other cases, when the photon beam is not narrow, or the

shield is relatively not thin, this exponential law is not valid, due to the

so called " build up" of photons in the point of interest. This build up

arises due to two modes of photon interaction with the matter, which

are: Compton scattering and pair production, while the photoelectric

effect does not yield any build up. Due to Compton scattering some

photons, which are emerged far away from the point of interest may be

scattered and as a result of this scattering it may reach the point of

interest (see the photon 1 in fig 8-1). Additionally, multiple Compton

scattering may arise due to the large thickness of the shield, increasing

the number of photons that may reach the point of interest (see the

photon 2 on fig. 8-1). In the pair production the energy may not be

transferred completely to the matter, since one or even the two photons,

resulting from the annihilation of the positron with one electron may

escape out of the matter, reaching the point of interest (the photon 3 on

fig 8-1).

Fig. (8-1)

- The build up factor B is defined as the ratio of the total

number of photons It, which arrive the point of interest directly Id

1

2

e+

3

9

4

from the source and due to scattering or pair production Is to the

number of photons, which arrive the same point directly Id, i.e:

B = It / Id =

= ( Is + Id ) / Id (8-6)

- The build up factor B strongly depends on photon energy E, as

well as on the atomic number Z of the shield and on the thickness of this

shield. Its magnitude may vary from 1 in an ideal geometry (i.e. when

the photon beam is very narrow and the shield thickness is very thin) to

some orders of magnitude for the practical conditions. This makes the

application of the relation (8-5) for calculation of the shield thickness

practically invalid for real conditions, since it will yield much less

thickness. For this reason, the build up factor should be taken into

consideration, in shield calculation. The correct equation that should be

used, to take into consideration the build up factor is:

Ė = Ė0 B e – μa x

(8-7)

- It should be mentioned that the thickness x , which is sufficient

to decrease the dose rate at the point of interest to a certain value, is

dependent on the activity of the source. Increase of the activity of the

gamma-ray source requires corresponding increase of the shield

thickness to reach the required dose rate outside the shield.

8-3 Shielding of the neutron sources:

- It has been mentioned in chapter 2 that the material with low

atomic number Z, especially hydrogen, are considered as the best

moderators for fast neutrons, since these neutrons (with energy higher

than 1 MeV) requires not more than 18 collision, in average, with the

hydrogen nuclei (protons) to moderate them to thermal neutrons with

energies of about 0.025 eV. In light materials, rich with hydrogen, such

as paraffin wax, plastics, water, and 0thers, the thickness which is

required to moderate or slow down the fast neutrons varies within 20-25

cm. So, the use of a layer with this thickness, of any of these materials,

(or any other equivalent light materials), will be sufficient to moderate

fast neutrons to thermal ones.

- One of the main principles used to shield the neutron sources is

to moderate fast neutrons, which are emitted from all neutron sources

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and generators, then to absorb moderated neutrons by other material

with a high probability for capture (absorption) of thermal and slow

neutrons, and finally, to attenuate the gamma radiation emitted during

the neutron capture.

- There are some elements, such as cadmium, boron and others,

which are characterized by a very high cross section for thermal and

slow neutron radiative capture (i.e. thermal neutron absorption with the

emission of gamma photon). So, after moderation of fast neutrons, a

layer of 1 mm thick of cadmium (Cd) is sufficient to absorb the majority

of the moderated neutrons.

- Due to the radiative capture of the thermal neutrons gamma ray

photons will be emitted, so that, an additional layer of a material with

high atomic number Z, such as lead (Pb) is required to attenuate these

radiations. For this reason, a third layer, with a reasonable thickness of

lead, is used to absorb gamma photons emitted from cadmium layer.

- So the ideal shield fort neutron sources consists of three

consequent layers which are: 20-25 cm wax, plastic or water, surrounded

by about 1 mm cadmium sheet, which in its turn, is surrounded by a

reasonable thickness of lead or any other high Z material.

- Neutron shields may be made by a single layer of a low Z

material, such as paraffin wax, plastic, water or others. This is related

with the limited ability of hydrogen and some other light (or low Z)

material to absorb thermal neutrons after their moderation. The main

requirement for such shields is to increase the thickness of the layer to a

sufficient value, so that the reduction of the dose rate, arising from the

neutron source outside the shield is achieved. Examples of such shields

are the neutron semi-spherical howitzers made of wax around the

neutron sources used for educational and other purposes. The thickness

of the paraffin wax or the water around the source may vary from about

50 to more than150 cm, depending on the neutron yield of the source.

- In case of accidents with neutron sources, one may use any

available light materials to shield the exposed neutron source, including

water bags, sands and rocks, and even pieces of wood.

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6

THE ORGANIZATIONAL ASPECTS

OF RADIATION PROTECTION

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GENERAL FRAMEWORK AND REQUIREMENTS

FOR RADSIATION PROTECTION

1. INTRODUCTION:

- Occupational exposure to radiation can occur as a result of

various human activities, including the use of radioactive sources and x-

ray machines in medicine, scientific research, agriculture and industry,

work associated with the different stages of the nuclear fuel cycle, and

occupations that involve the handling of materials containing enhanced

concentrations of naturally occurring radionuclides

- This lecture addresses the organizational aspects of radiation

protection, in situations of both normal and potential exposure. The

intention is to provide an integrated approach to the control of normal

and potential exposures due to external and internal irradiation from both

artificial and natural sources of radiation.

2. ADMINISTARATIVE REQUIREMENTS:

2-1 The practice and the intervention:

- The practices are defined as the human activities that add

radiation exposure to that which people normally receive from existing

radiation sources, or that increase the likelihood (i,e probability) of their

incurring exposure. For a practice, provisions for radiation protection

and safety can be made before its commencement, and the associated

radiation exposures and their likelihood can be restricted from the outset.

- The interventions are human activities that seek to reduce the

existing radiation exposure, or the likelihood of incurring exposure, and

which are not part of a controlled practice. In the case of intervention,

the circumstances giving rise to exposure or the likelihood of exposure

already exist, and their reduction can only be achieved by means of

protective or remedial actions.

2-2 The requirements of radiation protection:

a- The basic obligation:

- No practice shall be adopted, introduced, conducted,

discontinued or ceased and no source within a practice shall, as

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applicable, be mined, milled, processed, designed, manufactured,

constructed, assembled, acquired, imported, exported, distributed, sold,

loaned, hired, received, sited, located, commissioned, possessed, used,

operated, maintained, repaired, transferred, decommissioned,

disassembled, transported, stored or disposed of, except in accordance

with the national requirements, unless the exposure from such practice

or source is excluded from the requirements, including the requirements

of notification and authorization.

b- The notification

- Any legal person, who intends to carry out any of the actions

specified under the basic obligation for a practice or source shall submit

a notification to the Regulatory Authority of such an intention.

c- Authorization: registration or licensing

- The legal person responsible for any sealed source, unsealed

source or radiation generator (including x-ray machines, accelerators and

neutron generators) shall, unless the source is exempted, apply to the

Regulatory Authority for an authorization which shall take the form of

either a registration or a license.

- The legal person responsible for any irradiation installation, mine

or mill processing of the radioactive ores, installation processing

radioactive substances, nuclear installation or radioactive waste

management facility, or for any use of a source which the Regulatory

Authority has not designated as suitable for registration, shall apply to

the Regulatory Authority for an authorization which shall take the form

of a license.

- Any legal person applying for an authorization shall:

(a) submit to the Regulatory Authority and, if applicable, the

relevant information necessary to support the application;

(b) refrain from carrying out any of the actions described in

the basic obligation until the registration or license has been

granted;

(c) make an assessment of the nature, magnitude and

likelihood of the exposures attributed to the source and take all

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necessary steps for the protection and safety of both workers and

the public; and

(d) if the potential for an exposure is greater than any level

specified by the regulatory authority, have a safety assessment

made and submitted to the regulatory authority as part of the

application.

- The legal person responsible for a source to be used for medical

exposure shall include in the application for authorization:

(a) the qualifications in radiation protection of the medical

practitioners who are to be so designated by name in the

registration or license; or

(b) a statement that only medical practitioners with the

qualifications in radiation protection specified in the relevant

regulations or to be specified in the license will be permitted to

prescribe medical exposure by means of the authorized source.

- Licensee shall bear the responsibility for setting up and

implementing the technical and organizational measures that are needed

for ensuring protection and safety for the sources for which he is

authorized. He may appoint other people to carry out actions and tasks

related to these responsibilities, but He shall retain the responsibility for

the actions and tasks himself.

- licensee shall specifically identify the individuals responsible for

ensuring compliance with the national requirements.

- licensee shall notify the Regulatory Authority of his intentions to

introduce modifications to any practice or source for which he is

authorized, whenever the modifications could have implications for

protection or safety, and shall not carry out any such modification unless

specifically authorized by the Regulatory Authority.

d- Inspection:

- The Licensee shall permit duly authorized representatives of the

Regulatory Authority, and of the relevant Sponsoring Organizations

when applicable, to inspect their protection and safety records and to

carry out appropriate inspections of their authorized activities. Some

inspection should be announced and the others must not be announced.

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e- None-compliance

- In the event of a breach of any applicable requirement of the

regulations, the licensee shall investigate the breach and its causes,

circumstances and consequences, take appropriate action to remedy the

circumstances that led to the breach and to prevent a recurrence of

similar breaches, and communicate to the regulatory authority on the

causes of the breach and on the corrective or preventive actions taken or

to be taken.

- The communication of a breach of the regulations shall be

prompt and it shall be immediate whenever an emergency exposure

situation has developed or is developing. Failure to take corrective or

preventive actions within a reasonable time in accordance with national

regulations shall be grounds for modifying, suspending or withdrawing

any authorization that had been granted by the Regulatory Authority.

2-3 The basic framework of radiation protection:

- The principles of radiation protection and safety for are as

follows:

(a) Justification of practices:

“Any practice, or a source within the practice should not be

authorized unless this practice produces sufficient benefits to the

exposed individuals or to society to offset the radiation harm that it

might cause; that is: unless the practice is justified, taking into account

the social and economic factors.”

- The process of determining whether a practice is justified

involves consideration of all the radiation doses received by workers and

members of the public, for present and next generations

(b) Optimization of protection and safety:

- In relation to exposures from any particular source within a

practice, except for therapeutic medical exposures, protection and safety

shall be optimized in order that the magnitude of individual doses, the

number of people exposed and the likelihood of incurring exposures all

be kept as low as reasonably achievable, (ALARA principle), economic

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and social factors being taken into account, with the restriction that the

doses to individuals delivered by the source be subject to dose

constraints.

- Senior management should translate their commitment to

optimization of radiation protection into effective action by establishing

appropriate radiation protection program, and rules commensurate with

the level and the nature of the radiological risk presented by the practice.

- In order to reduce or avert exposures in intervention situations,

protective actions or remedial actions shall be undertaken whenever they

are justified.

(c) Dose limitation: “The normal exposure of individuals shall be restricted so that

neither the total effective dose nor the total equivalent dose to relevant

organs or tissues, caused by the possible combination of exposures from

authorized practices, exceeds any relevant dose limit, specified by the

national Regulatory Authority.

- The dose limits are applied, only, for occupational exposures and

general public exposures, while they are not applied for medical

exposures and exposures during emergencies.

- The limit on effective dose represents the level above which the

risk of stochastic effects due to radiation is considered to be

unacceptable, while it is much less than the thresholds for deterministic

effects. For localized exposure of the lens of the eye, extremities and the

skin, this limit on effective dose is not sufficient to ensure the avoidance

of deterministic effects, and therefore limits on equivalent dose are

specified for such situations.

(d) Guidance levels for medical exposure:

- Guidance levels for medical exposure shall be established for use

by medical practitioners. The guidance levels are intended to be a

reasonable indication of doses for average sized patients. They are

needed to provide guidance on what is achievable with current good

practice rather than on what should be considered optimum

performance;

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3- MANAGEMENT REQUIREMENTS FOR PROTECTION:

3-1 Safety culture:

- One of the definition of the safety culture is that it is consisting

of the assembly of characteristics and attitudes in organizations and

individuals which establishes that, as an overriding priority, protection

and safety issues receive the attention warranted by their significance

- A safety culture shall be fostered and maintained to encourage a

questioning and learning attitude to protection and safety and to

discourage complacency, which shall ensure that policies and procedures

be established that identify protection and safety as being of the highest

priority and problems affecting protection and safety be promptly

identified and corrected

3-2 Quality assurance:

- Quality assurance programs shall be established, that provide

adequate assurance that the specified requirements relating to protection

and safety are satisfied and quality control mechanisms for assessing the

effectiveness of protection and safety measures are fulfilled.

3-3 Human factors:

- Provision shall be made for reducing as far as practicable the

contribution of human error to accidents and other events that could give

rise to exposures, by ensuring that all personnel on whom protection and

safety depend be appropriately trained and qualified so that they

understand their responsibilities and perform their duties according to

defined procedures, and appropriate equipment, safety systems, and

procedural requirements be provided and other necessary provisions be

made to reduce, as far as practicable, the possibility that human error.

3-4 Qualified experts:

- Licensee should identify qualified experts and shall made

available the expertise for providing advice on the observance of the

regulations.

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4- THE PRINCIPAL REQUIREMENTS:

4-1 Security of sources:

- Sources shall be kept secure so as to prevent theft or damage and

to prevent any unauthorized legal person from carrying out any of the

actions specified in the basic obligation for practices, by ensuring that

control of the sources is ensured against lost or theft and any source

shall not be transferred unless the receiver possesses a valid

authorization. A periodic inventory of all sources, especially movable

shall be conducted at appropriate intervals to confirm that they are in

their assigned locations and are secure.

4-2 Defense in depth:

- A multilayer (defense in depth) system of provisions for

protection and safety, commensurate with the magnitude and likelihood

of the potential exposures involved, shall be applied to sources such that

a failure at one layer is compensated for or corrected by subsequent

layers, for the purposes of preventing accidents that may cause exposure

and mitigating the consequences of any such accident that does occur;

and restoring sources to safe conditions after any such accident.

4-3 Good engineering practice:

- As applicable, the sitting, location, design, construction,

assembly, commissioning, operation, maintenance and decommissioning

of sources within practices shall be based on sound engineering which

shall take account of approved codes and standards and other

documented instruments. This include taking into account of relevant

developments in technical criteria, as well as the results of any relevant

research on protection or safety and lessons from previous experiences.

5- VERIFICATION OF SAFETY:

5-1 Safety assessments:

- Safety assessments related to protection and safety measures for

sources within practices shall be made at different stages, including

sitting, design, manufacture, construction, assembly, commissioning,

operation, maintenance and decommissioning, as appropriate, in order to

identify the ways in which normal and potential exposures could be

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incurred, account being taken of the effect of events external to the

sources, as well as events directly involving the sources and their

associated equipment and to assess the quality and extent of the

protection and safety provisions.

5-2 Monitoring and verification of compliance:

- Monitoring and measurements shall be conducted of the

parameters necessary for verification of compliance with the

requirements and regulations. For this purpose, suitable equipment shall

be provided and verification procedures introduced. The equipment shall

be properly maintained and tested and shall be calibrated at appropriate

intervals with reference to standards traceable to national or international

standards.

5-3 Records:

- Different records shall be maintained for the practices and

sources and of the results of monitoring and verification of compliance,

including records of the tests and calibrations carried out in accordance

with the Standards.

6- CONDITION OF SERVICE:

6-1 Pregnant workers:

- A female worker should, on becoming aware that she is pregnant,

notify the employer in order that her working conditions may be

modified if necessary. The notification of pregnancy shall not be

considered a reason to exclude a female worker from work; however, the

employer of a female worker who has notified pregnancy shall adapt the

working conditions in respect of occupational exposure so as to ensure

that the embryo or foetus is afforded the same broad level of protection

as required for members of the public.

6-2 Conditions for young persons

- No person under the age of 16 years shall be subjected to

occupational exposure, and no person under the age of 18 years shall be

allowed to work in a controlled area unless supervised, and then only for

the purpose of training.

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6-3 Classification of areas

- The licensee shall designate controlled and supervised areas, in which

specific protective measures or safety provisions are required for

controlling normal exposures or preventing the spread of contamination

during normal working conditions. The licensee shall delineate

controlled and supervised areas by physical means.

- According to Saudi national regulation, the controlled area is

defined as the area in which the annual effective dose may reach 3/10

the occupational annual dose limit (i.e may reach 6 mSv/year).

- According to Saudi national regulation, the supervised area is

defined as the area in which the annual effective dose may reach 1/10

the occupational annual dose limit (i.e may reach 2 mSv/year).

6-4 The local rules and supervision:

- The licensee shall establish in writing local rules and procedures

as are necessary to ensure adequate levels of protection and safety for

workers and general public, and shall include in these rules the values of

any relevant investigation level or authorized level, and the procedure to

be followed in the event that any such value is exceeded.

- The licensee shall make local rules and the protective measures

and safety provisions known to those workers, to whom they apply and

to other persons who may be affected by them.

6-5 Personal protective equipment:

- The licensee shall ensure that workers are provided with suitable

and adequate personal protective equipment which meets any relevant

specifications, including protective clothing, protective respiratory

equipment, protective aprons and gloves, and organ shields.

- Workers must receive adequate instruction in the proper use of

protective equipment, especially respiratory equipment, including testing

for good fit. All personal protective equipment shall be maintained in

proper condition and tested at regular intervals.

- Tasks requiring the use of some specific personal protective

equipment shall be assigned, only to workers whom the basis of medical

advice, are capable of safely sustaining the extra effort necessary;

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RESPONSIBILITIES OF PARTIES

1- RESPONSIBLE PARTIES FOR RADIATION PROTECTION:

- The Regulatory Authority (and the intervening organizations in

the case of intervention) shall be responsible for the enforcement of the

regulations.

- The principal party having the main responsibilities for the

compliance with the regulations is he licensee (the employer).

- Other parties, who have subsidiary responsibilities for the

application of the regulations include:

(a) the radiation protection officers;

(b) the qualified experts;

(c) the medical practitioners;

(b) the workers;

(e) the suppliers;

(f) the Ethical Review Committees; and

(g) any other party to whom a principal party has delegated specific

responsibilities.

- All the parties shall have the general and specific responsibilities

set out in the national regulations.

2- RESPONSIBILITIES OF THE LICENSEE:

2-1 The general responsibilities:

- The general responsibilities of the principal party (the licensee),

within the requirements specified by the Regulatory Authority, are:

(a) to establish protection and safety objectives in conformity with

the requirements.

(b) to develop, implement and document a protection and safety

program, commensurate with the nature and extent of the risks

associated with the practices and interventions under his

responsibility, and sufficient to ensure compliance with the

requirements.

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(c) to determine the measures and resources needed to achieve the

protection and safety objectives and to ensure that the resources

are provided, and the measures are properly implemented.

(d) to keep the measures and resources continually under review, and

to verify that the protection and safety objectives are being

achieved.

(e) to identify any failures and shortcomings in the protection and

safety measures and resources, and to take steps to correct them

and prevent heir recurrence.

2-2 Specific responsibilities of the licensee:

- The licensee, who is engaged in activities involving normal or

potential exposure, shall appoint radiation protection officer (officers)

RPO, for carrying out the technical actions and tasks related to his

responsibilities, but the licensee shall retain the responsibility for these

actions.

- The licensee (employer of workers) shall be responsible for the

protection of workers from occupational exposure and for compliance

with the relevant requirements of the regulations.

- To fulfill his responsibilities, the licensees shall ensure, for all

workers engaged in activities that involve or could involve

occupational exposure, that:

(a) The occupational exposures are limited to the national limits.

(b) The occupational protection and safety are optimized in

accordance with the requirements.

(c) Decisions regarding measures for occupational protection and

safety are recorded and made available to the relevant parties.

(d) Policies, procedures and organizational arrangements for

protection and safety are established for implementing the

relevant requirements.

(e) Suitable and adequate facilities, equipment and services for

protection and safety are provided, the nature and extent of

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which are commensurate with the expected magnitude and

likelihood of the occupational exposure.

(f) Necessary health surveillance and health services are provided.

(g) Appropriate protective devices and monitoring equipment are

provided and arrangements are made for its proper use.

(h) Suitable and adequate human resources and appropriate training

in protection and safety are provided, as well as periodic

retraining and updating, as required, in order to ensure the

necessary level of competence

(i) Adequate records are maintained as required by the regulations

(j) Arrangements are made to facilitate consultation and co-

operation with workers with respect to protection and safety.

(k) Necessary conditions to promote a safety culture are provided.

2-3 Specific responsibilities of the RPO:

- The RPO is an individual, technically competent in radiation

protection matters, relevant to a given type of practice, who is qualified

through the judgment of the Regulatory Authority, and who is

designated by the licensee to oversee the implementation of the

requirements of the regulations.

- The responsibilities of the RPO

(a) Preparation of all technical aspects and procedures of radiation

protection program (including emergency plan and quality

assurance program) under the supervision of the qualified expert.

(b) Following-up of the implementation of the rules and procedures

for protection and safety, specified by the licensee, and

overseeing of the proper use of the surveying and monitoring

devices, the protective equipment, and all other equipments.

(c) Conduct all technical aspects, related with radiation protection

and safety, including different radiological surveys, monitoring,

tests and calibrations, or supervise their conduction at the

authorized parties.

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(d) Co-operate with the licensee and the regulatory authority with

respect to protection and safety and the operation of radiological

health surveillance and dose assessment programs.

(e) Register all technical data and personal dose information in the

records timely, and notify the workers who approach dose limit,

and discuss all protection affairs with workers and involved

parties.

(f) Conduct demonstrations and technical training on job concerning

radiation protection and safety of the authorized practices.

(g) Stop, promptly, any violation of the local rules or regulation, and

report to both the licensee and regulatory authority.

2-4 Specific responsibilities of the workers:

- Workers can by their own actions contribute to the protection and

safety. Workers shall:

(a) Follow the rules and procedures for protection and safety,

specified by the licensee;

(b) Use properly the monitoring devices, the protective equipment,

and clothing provided.

(c) Co-operate with the licensee with respect to protection and safety

and the operation of radiological health surveillance and dose

assessment programs.

(d) Provide to the licensee information on their past and current

work as is relevant to ensure effective and comprehensive

protection and safety for themselves and others;

(e) Abstain from any willful action that could put themselves or

others in situations that contravene the requirements of

regulations.

- Workers are also responsible for providing feedback to the

management, particularly when adverse circumstances arise

related to the radiation protection program.

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- Female workers have responsibilities regarding the protection of

the embryo or foetus. The worker herself “should, on becoming aware

that she is pregnant, notify the licensee in order that her working

conditions may be modified if necessary.

3. COOPRATION BETWEEN LICENSEES AND EMPLOYERS:

- The management of the occupational protection and safety of

transient, temporary or itinerant workers, and others who are employed

under contracts to organizations other than the operator, presents a major

concern. In order that these workers are adequately protected and do not

exceed any appropriate dose limit, there should be an adequate degree of

co-operation between the employer, the workers and the management of

the plants, for whom contracts are being undertaken.

- If workers are engaged in work that could involve a source that is

not under the control of their employer, the licensee responsible for the

source and the employer, shall co-operate by the exchange of

information to facilitate proper protective measures and safety

provisions of workers.

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3

NATIONAL (SA) DOSE LIMITS

FOR RADIATION EXPOSURES

1. THE TERMS "LIMIT" AND "LEVEL":

- The term "limit" is defined, for radiation protection purposes, as

the value of a quantity used in a certain specified activities or

circumstances, such as effective dose or equivalent dose, which must not

be exceeded.

- The term "level" is defined, for radiation protection purposes, as

the value of a specified quantity above which appropriate actions should

be considered:

- Among levels used in radiation protection, some most important

levels will be defined which are:

(a) The action level: is the level of dose rate or activity

concentration above which remedial actions or protective actions

should be carried out in chronic exposure or emergency exposure

situations, such as sheltering, immigration or others. Action

levels often serve to protect members of the public, but they also

have relevance in the context of occupational exposure in chronic

exposure situations, particularly that involving exposure to radon

in workplaces.

(b) The clearance level: is a value, established by the regulatory

authority, and expressed in terms of activity concentration or

total activity, below which sources of radiation may be released

from regulatory control.

(c) The guidance level for medical exposures: is the value of dose,

dose rate, or activity, selected by professional bodies in

consultation with the regulatory authority, to indicate a level

above which there should be a review by medical practitioners,

to determine wither or not the value is excessive, taking into

account the particular circumstances and applying sound clinical

judgment

(d) The intervention level: is the level of avertable dose at which a

specific protective action or remedial action is taken in an

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emergency exposure situation or chronic exposure situation. The

use of this term is normally confined to interventions related to

the protection of members of the public.

(e) The investigation level: The value of a quantity such as

effective dose, intake, or contamination per unit area or volume

at which an investigation should be conducted.

(f) The recording level: is a level of dose, exposure or intake

specified by the Regulatory Authority at which values of dose,

exposure or intake received by workers are to be entered in their

individual exposure records.

(g) The Reference level: is defined as a general term that can refer

to an action level, an intervention level, an investigation level or

a recording level. Such levels are helpful in the management of

operations as „trigger levels‟ above which some specified action

or decision should be taken.

2- RADIATION EXPOSURES:

- Radiation exposure is, generally, defined as the act or condition

of being subject to irradiation by ionizing radiation. The term exposure

is also used, in radiodosimetry, to express the amount of ionization,

produced in dry air by x-ray and low energy gamma radiation.

- In the general definition, exposure can be either external

exposure, when the irradiating source or sources are located outside the

body, or internal exposure when the source or sources are inside the

body (by inhalation, ingestion, injection or any other pathway of intake).

Moreover, exposure can be classified as:

(a) either normal or potential exposure.

(b) either occupational, medical or general public exposure.

(c) in intervention situation, either emergency or chronic exposure.

2-1 The normal exposure:

- The normal exposure is defined as an exposure which is expected

to be received under normal operating conditions of an installation or a

source, including possible minor mishaps that can be kept under control.

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2-2 The potential exposure:

- It is defined as the exposure that is not expected to be delivered

with certainty, but may result from an accident at a source or owing to

an event or sequence of events of a probabilistic nature, including

equipment failures and operating errors.

2-3 The occupational exposure:

- It is defined as all exposures of workers incurred in their work

due to this work, with the exception of exposures excluded from the

national regulations, and exposures from practices or sources exempted

by these regulations. The exposure of medical doctors, and other

technical and nursing staff belongs to occupational exposure.

2-4 The medical exposure:

- It is defined as exposure incurred by:

(a) patients as a part of their own medical or dental diagnosis or

treatment.

(b) exposures incurred by persons, other than those occupationally

exposed, knowingly while voluntarily helping in the support and

comfort of patients.

(c) exposures incurred by volunteers in a program of biomedical

research involving their exposure.

2-5 General public exposure:

- It is defined as exposure incurred by the members of the general

public from radiation sources, excluding any occupational or medical

exposure and the normal local natural background radiation, but

including exposure from authorized sources and practices and

intervention situations.

2-6 Chronic exposure:

- It is defined as exposure persisting in time and incurred with,

relatively, small dose rates.

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2-6 Acute exposure:

- It is defined as exposure incurred in a short time interval

(minutes, hours or days) with very high dose rate.

2-6 Emergency exposure:

- It is defined as exposure incurred in an emergency situation.

3- THE OCCUPATIONAL DOSE LIMITS:

3-1 Occupational limits for adult workers:

- The limits on effective dose for occupational exposure apply to

the sum of effective doses from external sources and committed

effective doses from intakes in the same period.

- The occupational exposure of any worker must not be exceeded

the following values:

(a) an effective dose of 20 mSv per year averaged over five

consecutive years.

(b) an effective dose of 50 mSv in any single year, provided that the

effective dose does not exceed 100 mSv over five consecutive

years.

(c) an equivalent dose to the lens of the eye of 150 mSv in a year

(d) an equivalent dose to the extremities (hands and feet) or the skin

of 500 mSv in a year.

3-2 Occupational limits for apprentices of 16- 18 years age:

- Separate limits are specified for apprentices of age 16–18 years,

who are training for employment involving exposure to radiation, and

for students of age 16–18 years, who need to use sources in the course of

their studies.

- The occupational exposure for this age category of trainees must

not exceed:

(a) an effective dose of 6 mSv in a year.

(b) an equivalent dose to the lens of the eye of 50 mSv in a year.

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(c) an equivalent dose to the extremities or the skin of 150 mSv in a

year.

3-3 Occupational limits for pregnant workers:

- The Occupational exposure for pregnant woman must not exceed

an effective committed dose of 1mSv in the tatal period of pregnancy,

when the irradiation is internal (i.e intake), and the effective dose must

not exceed 2 mSv, when the irradiation is external, during the 9 months

of pregnancy, in order to keep the effective dose to the embryo and

foetus not more than 1 mSv.

4- THE DOSE LIMITS FOR GENERAL PUBLIC:

- The estimated average doses to a relevant critical group that are

attributable to practices must not exceed the following limits:

(a) an effective dose of 1 mSv per year averaged over five

consecutive years.

(b) in a special circumstances, an effective dose of up to 2 mSv/year

provided that the effective dose does not exceed 5 mSv over five

consecutive years.

(c) an equivalent dose to the lens of the eye of 15 mSv in a year

(d) an equivalent dose to the extremities (hands and feet) or the skin

of 50 mSv in a year.

5- THE DOSE LIMITS FOR MEDICAL EXPOSURES:

- For medical exposure there are no limits. Instead of that the

principle of guidance levels is used.

- In medical exposure there is a dose limit for those persons, who

are offering comfort or supporting patients, undergoing medical

diagnosis or treatment, or visitors of such patients. The dose for these

comforters or visitors of patient shall be constrained to 5 mSv, during

the period of diagnosis or treatment of the patient.

- The dose to children visiting patients who have ingested

radioactive material shal be constrained to 1 mSv.

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6- THE DOSE LIMITS FOR EMERGENCY EXPOSURES:

- For emergency situations the objective should be to keep doses to

intervening personnel below an effective dose of 100 mSv or equivalent

doses of 1 Sv to the skin and 300 mSv to the lens of the eye in some

situations.

- However, where life saving actions are concerned significantly

higher levels of dose could be justified, although every effort should be

made to keep doses below ten times the maximum single year dose limit

(i.e. below 500 mSv) in order to avoid deterministic effects on health.

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THE RADIATION PROTECTION PROGRAM (RPP)

1. INTRODUCTION

- It has been mentioned that the licensee shall establish a radiation

protection program (RPP), which is one of the requirements for all

phases of a practice, and to the lifetime of a facility, from design through

process control to decommissioning. The general objective of RPP is to

reflect the application of the management responsibility for radiation

protection and safety through the adoption of management structures,

policies, procedures and organizational arrangements that are

commensurate with the nature and extent of the risks.

- Prior to establishment of the RPP for a practice, a radiological

evaluation shall be conducted to describe, as precisely as necessary, the

situation involving occupational, medical and public exposures. This

evaluation should include, all aspects of operations an identification of

the sources of routine and potential exposures and a realistic estimate of

the relevant doses and probabilities

- the legal person (licensee) applying for an authorization should

make an assessment of the nature, magnitude and likelihood of the

exposures and, if necessary, a safety assessment. Such a safety

assessment should contribute to the design of the RPP.

2- STRUCTURE OF THE RPP

- The RPP covers the main elements contributing to protection and

safety of the practices and is therefore a key factor for the development

of protection and safety.

- The RPP is composed of 6 main elements (or components),

which are:

(a) A committed administration to safety and protection.

(b) Selection of personnel and their training.

(c) An effective surveillance for occupational exposure.

(d) An effective surveillance for public exposure.

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(e) A proper quality assurance program.

(f) An emergency response plan.

2-1. A COMMITTED ADMINISTRATION AND ASSIGNMENT

OF RESPONSIBILITIES

- The highest managerial level should submit, in written, the

policy statements which ensure that radiation protection in the practices,

related with radiation exposure, deserves the highest consideration at all

levels. The licensee shall appoint other people to carry out actions and

tasks related to responsibilities in radiation protection aspects, but he

shall retain the responsibility for the actions and tasks himself. The

licensee shall specifically identify the individuals responsible for

ensuring compliance with the national regulations. The responsibilities

of each level, from the top management to the workers, regarding each

aspect of the RPP should be clearly delineated and documented in

written policy statements to ensure that all are aware of them. Radiation

protection officer (or officers) must be appointed, to oversee the

application of the regulatory requirements.

- The organizational structures at the licensee should reflect the

assignment of responsibilities, and the commitment of the organization

to protection and safety. The management structure should facilitate co-

operation between the various individuals involved. The RPP should be

designed in such a way that the relevant information is provided to the .

2-2. SELECTION OF PERSONNEL AND THEIR TRAINING:

- Criteria for selection of personnel should be defined, including

medical and moral aspects, and technical educational levels

- It may be appropriate, depending on the size of the organization,

to create a specific committee with representatives of those departments

concerned with radiation exposures. The main role of this committee

would be to advise senior management on the RPP individuals in charge

of the various aspects of the work.

- Qualified experts in radiation protection should be identified and

made available for providing advice on the observance of the Standards.

- Senior management should be trained in the risks associated with

ionizing radiation, the basic principles of radiological protection, their

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main responsibilities regarding radiation risk management and the

principal elements of the RPP.

- Training for those workers directly involved in work with

radiation sources should include relevant information, presented in the

form of documents, lectures and applied training, which emphasizes

procedures specific to the worker‟s job assignment.

- Training programs should be documented and approved at an

appropriate level within the organization. Such programs should be

reviewed periodically to ensure that they remain up to date.

2-3. AN EFFECTIVE SURVEILLANCE FOR OCCUPATIONAL

EXPOSURE: - An effective surveillance for occupational radiation protection

shall be established. This surveillance shall include:

- Systems and procedures for securing radioactive source (or

sources) and an accountability system, which includes records of these

sources.

- A health surveillance program should be prepared and health

criteria should be established for radiation workers..

- Classification of working areas, whenever there is occupational

exposure to radiation. These areas should be clearly defined as part of

the RPP, and their classification should result from the prior radiological

evaluation. The two types of areas, which are controlled and supervised,

shall be delineated. Restriction of access to the controlled and supervised

by permits, and physical barriers, locks or interlocks shall be provided.

- Establishment of occupational radiation protection and safety

measures, including rules and procedures that are appropriate,

- Local rules, describing the organizational structures and the

procedures to be followed in classified areas, should be developed by

management and written down. The rules should be prominently

displayed or readily available in the workplace. and they should include

procedures and values of any relevant investigation level or authorized

level, and the procedure to be followed in the event that any such value

is exceeded. The local rules and procedures and the protective measures

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- A radiological monitoring program for workplaces shall be

established and implemented. The nature and extent of the monitoring

program shall provide the primary justification for radiological

protection Monitoring program can be divided and subdivided into a

number of different types. The first division relates to the objectives of

the monitoring. At this level, three types of monitoring are conducted for

radiation protection purposes. These are routine monitoring, which is

associated with continuing operations and is intended to demonstrate

that the working conditions, including the levels of individual dose,

remain satisfactory, and to meet regulatory requirements, task related

monitoring which applies to a specific operatio, and special monitoring

which is investigative in nature and typically covers a situation in the

workplace for which insufficient information is available to demonstrate

adequate control.

- Individual monitoring for internal or external dose assassment

shall be undertaken for any worker who is regularly employed in a

controlled or supervised areas or who enters a controlled area only

occasionally. Individual monitoring in a supervised area shall not be

required but the occupational exposure of the worker shall be assessed

This assessment shall be on the basis of the results of monitoring of the

workplace or individual monitoring. The nature, frequency and precision

of individual monitoring shall be determined with consideration of the

magnitude and possible fluctuations of exposure levels and the

likelihood and magnitude of potential exposures.

- To secure the necessary accuracy and precision, individual

dosimetry should be performed, whenever possible, by an approved

dosimetry service. The regulatory authority should give consideration to

the establishment of a national accreditation procedure as a basis for the

approval of dosimetry services.

- Record keeping is an essential part of the individual monitoring

process. In making records of dose assessments it is important to

establish the

- Many of records, for example the full details of a particular

radiation survey, are transitory in nature are only relevant for the lifetime

of an established review period, and there may be no need to retain such

records for extended periods. Other records may be related to decisions

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about the definition of the workplace, and these records may be relevant

for the lifetime of the workplace.

- Exposure records for each worker shall be preserved during the

worker‟s working life and afterwards at least until the worker attains or

would have attained the age of 75 years, and for not less than 30 years

after the termination of the work involving occupational exposure.

individual or if required by regulation.

2-4. AN EFFECTIVE SURVEILLANCE FOR PUBLIC

EXPOSURE: - An effective surveillance for radiation protection of general

public and the environment shall be established. This surveillance shall

include:

- Establishment of the efficient systems that will ensure securing

of the radioactive sources against accessing them for any unauthorized

person, and against loss and theft.

- Establishment of an effective surveillance for conduction of all

necessary radiological surveys and monitoring in all places, accessed by

the general public around the controlled and supervised areas, either

locally or by contracting with an authorized party.

- A radiological monitoring program for areas, which may be

affected from the licensed sources or authorized releases of

radionuclides, shall be established and implemented. The nature and

extent of the monitoring program shall provide the primary justification

for radiological protection for general public. This monitoring should

include conduction of environmental radiological monitoring in these

areas, by studying samples taken from these areas and their radiological

analyses to assess any environmental hazards.

- Establishment of safety measures for radioactive releases to the

environment, including rules and procedures that are appropriate,

- Optimization of the generation of radioactive waste, as low as

reasonably achievable, and establishment of measures, criteria and

procedure for safe interim storage of the generated radioactive waste

from the authorized practices and for safe disposal of that waste, in

accordance with the national regulations for waste disposal.

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- Record keeping, of all information related with all radioactive

releases, waste generation, and waste disposal of used sources,

radionuclides, or wastes.

- Establishment of measures, rules and procedures, that should be

implemented for safe transport of the radioactive material, and record

keeping for all transport process, so that to minimize the general public

exposures from the transport activities.

2-5 THE QUALITY ASSURANCE PROGRAM:

- A quality assurance (QA) program shall be established as part of

the RPP. The licensee shall be responsible for establishing the quality

assurance program required by the principal requirements of the national

regulation, and the nature and extent of the quality assurance program

shall be commensurate with the magnitude and the likelihood of the

potential exposures from the sources for which they are responsible

- The quality assurance program shall provide for planned and

systematic actions, aimed at providing adequate confidence that the

specified design and operational requirements related to protection and

safety are satisfied, including provisions for feedback of operational

experience. Additionally, it shall provide for validation of designs, and

supply and use of materials, of manufacturing, inspection and testing

methods, and of operating procedures.

- Maintaining the effectiveness of any RPP relies on the ability of

those in charge of implementing its various components to adopt a QA

program and to pay as much attention as possible to lessons learned from

experience. The evaluations through appropriate reviews and audits, of

the way in which the RPP is implemented are key elements of an

effective quality assurance program.

- Management should be committed to QA and should provide the

financial and human resources necessary to achieve quality standards

and to maintain them continuously.

- The RPP should be assessed on a regular basis. Audits and

reviews of activities within the RPP should be scheduled on the basis of

the status and importance of the activity. Management should establish a

process for such assessments to identify and correct administrative and

management problems that may prevent the achievement of program

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objectives. Audits and reviews should be conducted by persons who are

technically competent to evaluate the processes and procedures being

assessed, but do not have any direct responsibility for those activities.

- Audits and reviews should be performed in accordance with

written procedures and checklists.

2-6 THE EMERGENCY PLANNING:

- The licensee, responsible for sources for which prompt

intervention may be required, shall ensure that an emergency plan exists

that defines on-site responsibilities and takes account of off-site

responsibilities appropriate for the source and provides for

implementation of each form of protective action.

- The emergency or contingency plans should specify how the

responsibilities for the management of interventions will be discharged

on the site, off the site and across national boundaries.

- The emergency and contingency plans should include objectives

of the plans, scenarios for all possible accidents and incidents with the

authorized practices and sources and actions that should be taken in each

scenario.

- The emergency plans prepared in advance should include

definition of the roles and responsibilities of all workers concerned in

the emergency response. Details of protective actions to be taken,

protective clothing and monitoring instruments to be used, and

dosimetry arrangements should also be specified.

- The dose limits for workers should be assumed to apply unless

there are overriding reasons not to apply them. However, exceeding the

dose limit of exposure in an emergency situation may be permitted,

exclusively, for volunteers who know how to act correctly in the

prevailing situation. There are three situations where it would be

justified for the dose limits to be exceeded, which are

(a) for the purpose of saving life or preventing serious injuries.

(b) if actions intended to avert a large collective dose or to

prevent the development of catastrophic conditions.

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- An emergency team should be formed, and this team should be

trained on actions hat should be undertaken in different scenarios.

Additional training should be provided on use of protective clothing,

respiratory protective equipments, the means of shielding, and iodine

prophylaxis. Where workers may be exposed to radiation fields with

relatively high dose rates, pre-established guidance should be given on

dose, dose rates and air concentrations for the appropriate time period.

- Doses incurred by workers during the emergency phase of the

intervention should be recorded separately, if possible, from the doses

incurred during routine work, but should be noted on the workers‟ dose

records.

- In accordance with the conditions of authorization, management

should draw up formal plans to deal with situations in which workers

might be overexposed. These plans should address the management of

overexposed workers and the health consequences that might be

encountered. They should specify the necessary actions to be taken, and

management should allocate resources for carrying out those actions.

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SAFE TRANSPORT OF RADIOACTIVE MATERIAL

1. INTRODUCTION

- The transport regulations establish standards of safety which

provide an acceptable level of protection against radiation and thermal

hazards to persons, property and the environment that are associated

with the transport of radioactive material. This protection is achieved by

requiring:

(a) proper containment of the radioactive contents.

(b) control of the external radiation levels

(d) prevention of damage caused by heat.

- These requirements are satisfied firstly, by applying a graded

approach to contents limits for packages and conveyances, to the

performance standards applied to package designs depending upon the

hazard of the radioactive contents. Secondly, they are satisfied by

imposing requirements on the design and operation of packages and on

the maintenance of packaging, including a consideration of the nature of

the radioactive contents. Finally, they are satisfied by requiring

administrative controls including, approval by regulatory authorities.

2. DEFINITIONS

- For safe transport of radioactive material, special definitions are

applied. Some of these definitions are:

2-1 A1 and A2

- A1 means the activity value of a special form radioactive material

which is used to determine the activity limits in a type A package.

- A2 means the activity value of a radioactive material, other than

special form radioactive material, which is used to determine the activity

limits in a type A package.

2-2 Approval

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- Unilateral approval means an approval of a design which is

required to be given by the regulatory authority of the country of origin

of the design only

- Multilateral approval means approval by the regulatory authority

of the country of origin of the design or shipment and also, where the

consignment is to be transported through or into any other country,

approval by the regulatory authority of that country. The term “through

or into” specifically excludes “over”, i.e. the approval shall not apply to

a country over which radioactive material is carried in an aircraft,

provided that there is no scheduled stop in that country.

2-3 Carrier

- Carrier means any person, organization or undertaking the

carriage of radioactive material by any means of transport. The term

includes both carriers for hire or reward (known as common or contract

carriers in some countries) and carriers on own account (known as

private carriers in some countries).

2-4 Consignee

- Consignee means any person, organization or government which

receives a consignment.

2-5 Consignment

- Consignment means any package or packages, or load of

radioactive material, presented by a consignor for transport.

2-6 Consignor

- Consignor means any person, organization or government which

prepares a consignment for transport.

2-7 Contamination

- Contamination means the presence of a radioactive substance on

a surface in quantities in excess of 0.4 Bq/cm2 for beta and gamma

emitters and low toxicity alpha emitters, or 0.04 Bq/cm2 for all other

alpha emitters.

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- Non-fixed contamination means contamination that can be

removed from a surface during routine conditions of transport.

- Fixed contamination means contamination other than non-fixed

contamination.

2-8 Conveyance

- Conveyance means:

(a) for transport by road or rail: any vehicle,

(b) for transport by water: any vessel, or any hold,

compartment, or defined deck area of a vessel, and

(c) for transport by air: any aircraft.

2-9 Exclusive use

- Exclusive use means the sole use, by a single consignor, of a

conveyance or of a large freight container, in respect of which all initial,

intermediate and final loading and unloading is carried out in accordance

with the directions of the consignor or consignee.

2-10 Low dispersible radioactive material

- Low dispersible radioactive material means either a solid

radioactive material or a solid radioactive material in a sealed capsule,

that has limited dispersibility and is not in powder form.

2-11 Low specific activity material

- Low specific activity (LSA) material means radioactive material

which by its nature has a limited specific activity, or radioactive material

for which limits of estimated average specific activity apply. External

shielding materials surrounding the LSA material shall not be considered

in determining the estimated average specific activity. LSA material

shall be in one of three groups:

(a) LSA-I (b) LSA-II (c) LSA-III

2-12 Overpack

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- Overpack means an enclosure such as a box or bag, used by a

single consignor to facilitate as a handling unit a consignment of one or

more packages for convenience of handling, stowage and carriage.

2-13 Package

- Package means the packaging with its radioactive contents as

presented for transport. The types of packages which are subject to the

activity limits and material restrictions are:

(a) Excepted package;

(b) Type A package;

(c) Type B(U) package;

(d) Type B(M) package;

(e) Type C package.

2-14 Packaging

- Packaging means the assembly of components necessary to

enclose the radioactive contents completely. It may, in particular, consist

of one or more receptacles, absorbent materials, spacing structures,

radiation shielding and service equipment for filling, emptying, venting

and pressure relief; devices for cooling, or absorbing mechanical shocks.

2-15 Radiation level

- Radiation level means the corresponding dose rate expressed in

millisieverts per hour.

2-16 Shipment

- Shipment means the specific movement of a consignment from

origin to destination.

2-17 Special arrangement

- Special arrangement means those provisions, approved by the

regulatory authority, under which consignments which do not satisfy all

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the applicable requirements of the transport regulations may be

transported.

2-18 Special form radioactive material

- Special form radioactive material means either an indispersible

solid radioactive material or a sealed capsule containing radioactive

material.

2-19 Surface contaminated object

- Surface contaminated object (SCO) means a solid object which is

not itself radioactive but which has radioactive material distributed on its

surfaces. SCO shall be in one of two groups:

(a) SCO-I (b) SCO-II:

2-20 Transport index

- Transport index (TI) assigned to a package, overpack or freight

container means a number which is used to provide control over

radiation exposure.

3. GENERAL PROVISIONS

3-1 Radiation protection

- Doses to persons shall be below the relevant dose limits.

Protection and safety shall be optimized in order that the magnitude of

individual doses, the number of persons exposed, and the likelihood of

incurring exposure shall be kept as low as reasonably achievable,

economic and social factors being taken into account.

- A Radiation Protection Program shall be established for the

transport of radioactive material. The nature and extent of the measures

to be employed in the program shall be related to the magnitude and

likelihood of radiation exposures.

- A radioactive material shall be segregated from the transport

workers and from members of the public. For the purpose of calculation

of segregation distance, the following values should be used:

(a) For workers a dose of 5 mSv/year

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(b) For members of the public 1 mSv/year.

- Radioactive material shall be segregated from undeveloped

photographic films. The basis for determining the segregation distances,

is that the dose to these films shall not exceed 0.1 mSv for the whole

transport period.

- A package shall not contain any items other than those that are

necessary for the use of the radioactive material. The interaction

between these items and the package, under the conditions of transport

applicable to the design, shall not reduce the safety of the package.

- Tanks and intermediate bulk containers used for the transport of

radioactive material shall not be used for the storage or transport of other

goods unless decontaminated to below the level of 0.4 Bq/cm2 for beta

and gamma emitters and low toxicity alpha emitters and 0.04 Bq/cm2 for

all other alpha emitters.

- The transport of other goods with consignments being

transported under exclusive use shall be permitted provided the

arrangements are controlled only by the consignor and it is not

prohibited by other regulations.

- Consignments shall be segregated from other dangerous goods

during transport, in compliance with the transport regulations for these

goods.

3-2 Control of contamination and leaking packages

- The non-fixed contamination on the external surfaces of any

package shall be kept as low as practicable and, under routine conditions

of transport, shall not exceed the following limits:

(a) 4 Bq/cm2 for beta and gamma emitters and low toxicity alpha

emitters, and

(b) 0.4 Bq/cm2 for all other alpha emitters.

These limits are applicable when averaged over any area of 300

cm2 of any part of the surface.

- If it is evident that a package is damaged or leaking, or if it is

suspected that the package may have leaked or been damaged, access to

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the package shall be restricted and a qualified person shall, as soon as

possible, assess the extent of contamination and the resultant radiation

level of the package. The scope of the assessment shall include the

package, the conveyance, the adjacent loading and unloading areas, and,

if necessary, all other material which has been carried

4. DETERMINATION OF THE TRANSPORT INDEX

4-1 Determination of the TI:

- The transport index (TI) for a package, overpack or freight

container, or for unpackaged LSA-I or SCO-I, shall be the number

derived in accordance with the following procedure:

(a) Determine the maximum radiation level (experimentally) in

units of millirem per hour (mrem/h) at a distance of 1 m from the

external surfaces of the package, overpack, freight container. The

resulting number is the transport index.

For uranium and thorium ores and their concentrates, the

maximum radiation level at any point 1 m from the external

surface of the load may be taken as:

(i) 40 mrem/h for ores and physical concentrates of ores

(ii) 30 mrem/h for chemical concentrates of thorium;

(iii) 2 mrem/h for chemical concentrates of uranium, other than

uranium hexafluoride.

(b) For tanks, freight containers and unpackaged LSA-I and

SCO-I, the value determined in step (a) above shall be multiplied

by the appropriate factor from Table 1.

Table 1: Multiplication factors for large dimension load

Size (i.e area) of load Multiplication factor

1 m2 1

1 m2 < area 5 m

2 2

5 m2 < area 20 m

2 3

20 m2 < area 10

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(c) The value obtained in steps (a) and (b) above shall be rounded

up to the first decimal place (e.g. 1.13 becomes 1.2), except that

a value of 0.05 or less may be considered as zero.

- The transport index for each overpack, freight container or

conveyance shall be determined as either the sum of the TIs of all the

packages contained, or by direct measurement of radiation level, except

in the case of non-rigid overpacks, for which the transport index shall be

determined only as the sum of the TIs of all the packages.

4-2 Limits on the TI and radiation level:

- Except for consignments under exclusive use, the transport index

of any package or overpack shall not exceed 10.

- Except for consignments transported under exclusive use or

special arrangement the maximum radiation level at any point on the

external surface of a package or overpack shall not exceed 2 mSv/h.

- The maximum radiation level at any point on the external surface

of a package or overpack, under exclusive, use shall not exceed 10

mSv/h.

5. CATEGORIES OF PACKAGES:

- Packages and overpacks shall be assigned to either category

WHITE-I, YELLOW-II or YELLOW-III in accordance with the

conditions specified in Table 2 and with the following requirements:

(a) For a package or overpack, both the transport index and

the surface radiation level conditions shall be taken into account

in determining the appropriate category for it. Where the

transport index satisfies the condition for one category but the

surface radiation level satisfies the condition for a different

category, the package or overpack shall be assigned to the higher

category. For this purpose, category WHITE-I shall be regarded

as the lowest category.

(b) If the surface radiation level is greater than 2 mSv/h, the

package or overpack shall be transported under exclusive use.

(c) A package transported under a special arrangement shall be

assigned to category YELLOW-III.

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Table 2: Categories of packages and overpacks

Conditions

Category Transport Index Maximum radiation level at

any point on the surface

0 Not more than 0.005 mSv/h WHITE -I

0 < TI 1 More than 0.005 mSv/h, but

not more than 0.5 mSv/h

YELLOW-II

1 < TI 10 More than 0.5 mSv/h, but not

more than 2 mSv/h

YELLOW-III

10 < TI More than 2 mSv/h YELLOW-III under exclusive use

6. MARKING AND LABELLING:

6-1 Marking:

- Each package shall be legibly and durably marked on the outside

of the packaging with an identification of either the consignor or

consignee, or both.

6-2 Labelling:

- Each package, overpack and freight container shall bear the

labels which conform to the models in Fig. 1, Fig. 2 or Fig. 4, except as

allowed under the alternative provisions for large freight containers and

tanks, according to the appropriate category. Any labels which do not

relate to the contents shall be removed or covered.

- The labels conforming to the models in Fig. 1, Fig. 2 and Fig. 3

shall be affixed to two opposite sides of the outside of a package or

overpack or on the outside of all four sides of a freight container or tank.

7. STORAGE IN TRANSIT:

7-1 Segregation during transport and storage in transit:

- Packages, overpacks and freight containers containing

radioactive material and unpackaged radioactive material shall be

segregated during transport and during storage in transit:

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(a) from workers in regularly occupied working areas by

distances calculated using a dose criterion of 5 mSv in a year and

conservative model parameters.

(b) from members of the critical group of the public, in areas

where the public has regular access, by distances calculated using

a dose criterion of 1 mSv in a year and conservative model

parameters;

(c) from undeveloped photographic film by distances calculated

using aradiation exposure criterion for undeveloped photographic

film due to the transport of radioactive material of 0.1 mSv per

consignment of such film; and

(d) from other dangerous goods.

- Category II-YELLOW or III-YELLOW packages or overpacks

shall not be carried in compartments occupied by passengers, except

those exclusively reserved for couriers specially authorized to

accompany such packages or overpacks.

7-2 Stowage during transport and storage in transit:

- Consignments shall be securely stowed. Provided that its average

surface heat flux does not exceed 15 W/m2 and that the immediately

surrounding cargo is not in sacks or bags, a package or overpack may be

carried or stored among packaged general cargo without any special

stowage provisions except as may be specifically required by the

regulatory authority in an applicable approval certificate.

- Loading of freight containers and accumulation of packages,

overpacks and freight containers shall be controlled as follows:

(a) Except under the condition of exclusive use, and for

consignments of LSA-I material, the total number of packages,

overpacks and freight containers aboard a single conveyance

shall be so limited that the total sum of the transport indexes

aboard the conveyance does not exceed the values 50. The same

rule is applied for storage in transit. For this purpose packages

are grouped in separate groups each with TI not exceeding 50.

(b) The radiation level under routine conditions of transport shall

not exceed 2 mSv/h, at any point on the external surface, and 0.1

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mSv/h at 2 m from the external surface of the conveyance,

except for consignments transported under exclusive use, by road

or railways.

- Any package or overpack having a transport index greater than

10 shall be transported, only, under exclusive use.

- groups of packages shall be stored so as to maintain a spacing of

at least 6 m between any two groups.

Fig. 1: Label for the category WHITE- I

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Fig. 2: Label for the category YELLOW-II

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Fig. 3: Label for the category YELLOW-III

a