Training Course
on
SAFETY ASPECTS IN RESEARCH APPLICATIONS
OF
IONISING RADIATIONS
Lecture Notes
ORGANISED BY
INDIAN ASSOCIATION FOR
RADIATION PROTECTION
IN COLLEBORATION WITH
Radiological Physics & Advisory Division
Bhabha Atomic Research Centre
CT&CRS, Anushaktinagar,
Mumbai – 400 094.
8.1
Training Course
on
SAFETY ASPECTS IN RESEARCH APPLICATIONS
OF
IONISING RADIATIONS
Lecture Notes
Compiled by P.K.Gaur
ORGANISED BY
INDIAN ASSOCIATION FOR
RADIATION PROTECTION
IN COLLEBORATION WITH
Radiological Physics & Advisory Division
Bhabha Atomic Research Centre
CT&CRS, Anushaktinagar,
Mumbai – 400 094.
8.2
Contents
1. Radiation Physics
2. Interaction of Ionising Radiation with Matter
3. Radiation Quantities & Units
4. Principles of Radiation Detection and Monitoring Devices
5. Biological Effects of Ionising Radiation
6. Operational Limits
7. Radiation Hazards Evaluation and Control
8. Planning of Radioisotope Laboratories
9. Regulatory Aspects of Radiological Safety
10.Disposal of Radioactive Waste
11.Transportation of Radioisotopes
12.Production of Radioisotopes and Labelled Compounds
Appendix-1 : Procedure for Establishing a Radioisotope Laboratory and
for the Procurement of Radio nuclides
Appendix-2 : Radiation Protection Survey of a Radioisotope Laboratory
Appendix-3 : Contamination Measurement & Decontamination Procedures
Appendix-4 : Calibration of Radiation Monitors
8.3
1. RADIATION PHYSICS
ATOMIC STRUCTURE
Matter is built up of individual entities called elements like hydrogen, oxygen,
carbon, etc. For example, water is made up of hydrogen and oxygen. An atom is the basic
component of each element. All atoms of a given element are identical, for instance, all
hydrogen atoms have the same properties.
An atom has a nucleus at its center. It is positively charged. The nucleus of
an atom is orbited by electrons, which are negatively charges. Electrons occupy one
or more orbits around the nucleus. The atomic nucleus is built up primarily of protons
(+vely charged particles) and neutrons (which carry no electric charge but have
almost the same mass as the protons). The positive charge of a proton is equal in
magnitude to the negative charge of an electron. The number of electrons orbiting around
the nucleus is always equal to the number of protons in the nucleus. Hence, the
atom is electrically neutral. The structure of an atom may be visualized as something
similar to that of our solar system with the sun at the center and the planets representing
the electrons orbiting around it.
An atom is so tiny that it cannot be seen
by
naked
eye
or
even
under
a
microscope. Its diameter is of the order
of 10-8 cm and that of the nucleus is of
the order of 10-12 - 10-13 cm. Since the
electrons have negligible mass, the mass
of an atom is concentrated in the nucleus
neutrons
with contributions entirely from protons
and neutrons. The mass of a proton and a
neutron are 1.6723 x 10-24gm and
electrons
protons
1.67747 x 10-24gm respectively. There
are 92 elements in nature. These
Fig. 1.1 : Oxygen Atom
were formed at the time earth was
formed. They progressively contain an
8.4
increasing number of protons, neutrons, and electrons. Hydrogen is the simplest of the
elements. It has only one proton (and no neutron) in the nucleus and one electron in its
orbit. Hydrogen is followed by helium, lithium, beryllium, boron, etc.
The number of protons in a nucleus is called the atomic number ‘Z’ of
an element which determines its place in the periodic table and its chemical properties.
The total number of protons and neutrons in the nucleus constitutes the
mass number ‘A’ of the atom and it approximates its atomic weight.
An element is specified by using its chemical symbol, the mass number, and
atomic number,
A
Z
X
: For example
1
1
H
(Hydrogen),
6
3
Li (Lithium)
This indicates that hydrogen atom has only one proton in its nucleus and no
neutron and lithium has three protons and three neutrons.
ISOTOPES
All atoms of the same element have the same number of protons and electrons and
thus the same atomic number. They can, however, have different numbers of neutrons
and are then called isotopes of that element. Thus isotopes of an element have the same
atomic number, but different mass numbers. Isotopes of an element are chemically
identical.
1
2
3
Example: 1 H , 1 H ,1 H
*
Isotopes of Hydrogen
59
27
60
Co, 27
Co *
*
Isotopes of Cobalt
radioactive isotopes
(radioisotope)
p
p
2n
p,n
Hydrogen
Deuterium
Fig. 1.2: Isotopes of Hydrogen Atom.
8.5
Tritium
RADIOACTIVITY
The stability of an element depends upon the ratio of neutrons to protons in
the nucleus. In elements of lower atomic number, except hydrogen, this ratio is generally
one. In heavier elements i.e., elements having higher atomic number, this ratio tends
to be more than unity, resulting in some degree of instability in the nucleus. An
unstable nucleus, which has excess energy, tries to attain stability by releasing this energy
in the form of radiation. Henry Becquerel, a French Physicist, discovered in 1996, that
a compound of uranium emitted some invisible radiant energy. Madame Curie studied
a large number of such substances and gave the name “Radioactivity” to this property.
Thus, radioactivity is a phenomenon in which an unstable nucleus of an element
disintegrates with the emission of energy and becomes a new element. It is a spontaneous
process unaffected by physical and chemical agents. Radioactivity exists in nature among
heavier elements. However, many lighter elements can be rendered radioactive by
bombardment with charged particles or neutrons. This is called artificial radioactivity.
Isotopes of an element having radioactive property are known as “Radioisotopes” (Refer
figure 1.2).
MODES OF RADIOACTIVE DISINTEGRATION
Radioactive transformation (or decay), whether natural or artificial can occur
only in a limited number of ways with the emission of 1) Alpha particles, 2) Beta
particles, 3) Positrons or by 4) Electron capture. However, the first two modes of
decay are the most commonly observed. It may also be noted that in a number of cases
the decay process is followed by the emission of gamma-rays or characteristic Xrays. Figure 1.3 shows schematically the various types of radiations emitted during
radioactive decay.
NATURE AND PROPERTIES OF NUCLEAR RADIATIONS
1. Alpha () Particles
An Alpha particle consists of 2 protons and 2 neutrons and it is identical to the
nucleus of a helium atom. Disintegration by alpha emission occurs only among higher
atomic number elements which are naturally radioactive. There are also a few artificially
8.6
produced alpha emitters. The resulting daughter nucleus after alpha decay has an atomic
number less by two and a mass number less by four than that of the parent nucleus. Alpha
emission may be followed by gamma emission. All alpha particles emitted from a
particular decay scheme have the same energy.
Thus alpha () decay from an element
A
Z
X  ZA42Y  24He
A
Z
X will be as follows:
e.g.
226
88
4
Ra  222
86 Rn  2 He
Some properties of alpha particles
1. They have a mass about four times that of the hydrogen nucleus and carry two
units of positive charge.
2. When alpha particles pass through matter, they cause ionization (process of
removing one or more electrons from the atoms present in the matter),
3. Alpha particles are less penetrating and they can be stopped even by a thin sheet
of paper.
Gamma
Rays
Beta particles
Alpha particles
Fig. 1.3: Types of Radiations Emitted during Radioactive Decay.
2. Beta particles ( -) and positrons (+)
These are electrons with negative and positive charge respectively. Since
electrons are not constituents of the nucleus, it is believed that they are created and
ejected from the nucleus in a radioactive transformation.
8.7
1
0
n11 p  10 e
(-) decay when n/p ratio is high in a nucleus
+ (position) emission is not as common as - emission. The daughter
product from - or + disintegration has the same mass number but a different
atomic number, with respect to the parent element.
Thus a -1 decay from an element
A
Z
X  Z A1W  10 e
A
Z
X will be as follows:
e.g.
60
27
60
Co 28
Ni   
Unlike alpha particles, all beta particles emitted from a given radioactive isotope
do not have the same energy but are distributed over a continuous energy spectrum.
There is a definite maximum energy associated with each beta emitting radioisotope and
is denoted by Emax.
Some properties of Beta Particles
1. Beta Particles are electrons with either negative (-) or positive charge (+positions).
2. They have a range of few meters in air depending upon their energy.
3. Like alpha particles, they ionize matter, but the ionization is less intense.
4. Beta particles, are comparatively, easily absorbed in matter, their penetrating
power depends mainly on their energy and on the atomic number of the absorber.
The emission of an alpha or beta particle is mostly accompanied by gamma rays.
If the nucleus retains some excess energy after the emission of a particle, it is said to be
in excited state and the excess energy is released almost immediately in the form of
gamma rays.
The slowing down of beta particles in matter (which is very effective in high Z
materials) results in the emission of electromagnetic radiation called bremsstrahlung. In
order to minimize the production of bremsstrahlung, it is better to use low Z shielding
materials, such aluminum or perspex, for beta radiation.
3. Electron Capture
This is an alternative process to positron emission. In this mode of decay, the
proton is changed into a neutron after capture of an electron, generally from the
8.8
innermost k-orbit. Here, the atomic number of the product nucleus is reduced by one unit
from that of the parent. As it will have one electron missing from the orbit, an electron
from the outer orbit will subsequently fall into the inner orbit accompanied by the
emission of radiation (X-rays) characteristic of the daughter atom.
4. Gamma Rays ()
The emission of an alpha or a beta particle is generally accompanied by the
emission of gamma rays, almost immediately. The daughter nucleus, mostly, has some
excess energy, after the emission of a particle and it is said to be in an excited state. The
excess energy is released in the form of gamma rays.
Some Properties of Gamma Rays
1. They are electromagnetic radiations, similar to X-rays and visible light.
2. They are highly penetrating. Higher the energy, higher is the penetrating power.
3. They do not possess any charge.
4. They ionize matter indirectly.
RADIOACTIVE DECAY
The decay of a radioactive sample is statistical in natural and it is impossible to
predict when a particular atom will disintegrate. The radioactivity in a sample decreases
constantly with time in a manner shown in figure 1.4 and follows an exponential law
which can be mathematically represented as
Ai  Ao .e   .t
Where A0 is the initial activity, At is the activity after a lapse of time ‘t’ and ‘’ is the
disintegrations or decay constant (fractional disintegration per unit time), which is a
characteristic of the radioisotope.
HALF-LIFE
In actual practice, the rate of decay of a radioisotope is usually in terms of its halflife, the time required for one-half of the atoms originally present to decay. Physical half
life of some of the radio nuclides used in research laboratories are given in table-1.1. The
half-life is usually denoted by t1/2 and is expressed by relation:
8.9
0.693

Radioactivity
t1 / 2 
Number of half lives
Fig. 1.4: Decay of Radioactivity with Time.
Half-life can be read directly from decay graphs as shown in figure 1.4. It may be
noted that in a time equal to one half-life, the radioactivity will be reduced by a factor of
2; at the end of second half-life, the activity will be reduced by a factor of 4 and so on.
Theoretically, it requires infinite time for the complete decay of a radioisotope, although
in most cases, a time of about 10 half-lives will reduce the activity to a negligible value
(1/1024) compared to the initial activity. Like, , the decay constant, half-life is also
characteristic of a radioisotope. This is because there is a relationship between  and halflife, given by the relation
 
0.693
t1 / 2
For different radioisotopes, the half-lives are different varying from fraction of a
second to several thousand years. Measurement of the half-life of a radioactive sample
8 . 10
helps in the identifications of the isotope. It should be noted that half-life of
radioactive isotope does not depend upon the amount of radioactivity.
Table 1.1: Some Common Radioisotopes used in Research Applications
Name
Amercium-241
Symbol
241
85
Krypton-85
Strontium-90Yttrium-90
Promethium-147
Thallium-204
Carbon-14
Phosphorous-32
Cobalt-60
Am
Kr
90
Sr- 90Y
147
Pm
Tl
14
C
32
P
60
Co
204
57
Cobalt-57
Iodine-125
Tritium-3
Caesium-137
Barium-133
Sulphur-35
Chromium-51
Calcium-45
Sodium-24
Co
125
I
3
H
137
Cs
133
Ba
35
S
51
Cr
45
Ca
24
Na
Sodium-22
22
Na
Type of Decay &
Energy in (MeV)
(5.48), (0.060)
Half-life
- (0.68), Brem
(051),
 (0.54, 2.27)
10.6 y
-
-
28 y
7.1 x 105
1.3 x 105
- (0.23)
- (0.77)
- (0.155)
- (1.71)
- (0.31), (1.17,
1.33)
Fe-X rays (0.122)
x-ray (0.027)
- (0.019)
 (0.514),  (0.66)
 (0.3, 0.082)
- (0.167)
e- (0.004),  (0.005)
- (0.257)
 (1.39),  (1037,
2.75)
+
 (0.545),  (0.511,
1.275)
2.6 y
3.76 y
5730 y
14 d
5.3 y
7.7 x 107
1.5 x 107
3.4 x 107
8.3 x 106
5.9 x 106
4.3 x 106
3.2 x 107
6.2 x 107
6.3 x 106
6.9 x 105
270 d
60 d
12.26 y
30 y
10 y
87.2 d
27.8 d
165 d
15 h
9.5 x 107
1.3 x 106
4.8 x 108
1.5 x 106
2.6 x 107
5.3 x 108
1.3 x 107
4.7 x 107
2.1 x 107
2.7 x 106
1.1 x 109
3.0 x 106
1.5 x 107
5.6 x 106
9.5 x 106
3.8 x 107
2.62 y
6.3 x 106
1.0 x 107
458 y
ALI (Bq)
Inhalation
Ingestion
1 x 105
5.1 x 102
y – years, d – days, h - hours
In the table the energy of beta particles is the maximum energy
(Emax) emitted.
---------
8 . 11
2. INTERACTION OF IONISING RADIATION WITH
MATTER
INTRODUCTION
Radiation emitted during the radioactive decay is characterized by their kinetic
energy. Their energies are expressed in units of electron volts (eV). An electron volt
corresponds to the kinetic energy acquired by an electron when it is accelerated through a
potential difference of 1 volt. Electron volt is a small unit. The energy of different
radiations emitted during radioactive decay (and X-rays) will be much higher and hence
expressed in terms of kilo (103) electron volts (keV) or million (106) electron
volts (MeV). In comparison, the energy of visible light is in the range of 1-4 eV.
Radiation emitted by a radioisotope cannot be seen or felt by any human senses.
However, when they are incident upon matter they interact with the atoms to produce
excitations and ionizations. Excitation is a process in which the orbital
electron of an atom is raised to a higher energy state, while in ionization one or
more electrons are removed from an atom, forming an ion pair, viz., a positive ion
and negative ion. Both these processes lead to transfer of energy from radiation to matter.
Because of their ability to ionize matter, these radiations are called ionising
radiations.
Ionisation is finally responsible for the observed biological, chemical, and
physical effects of radiation. They are also the means of detecting and measuring these
radiations. In this chapter, the interaction of ionizing radiation with matter is briefly
discussed.
Ionising radiations can be divided in to electromagnetic and particulate radiations.
Particulate radiations can be further divided in to charged and uncharged particles.
INTERACTION OF CHARGED PARTICLES
Alpha and beta are also known as charged particles. These charged particles lose
energy through collision with the electrons and nuclei of the atoms in the medium. This
leads to excitation and ionization along the path of a charged particle in the medium.
If the energy transferred to an electron in the medium is sufficient to remove it
completely out of the atom, the process is referred to as ionization. If the electron is just
8 . 12
raised to a higher energy state, the process is called excitation. Figure 2.1 shows
schematically the process of ionization by an alpha particle.
α- particle
Electron
α particle
Fig 2.1: Ionisation by alpha particles
The number of ion pairs produced per unit path length of charged
particle is called specific ionization. The specific ionization is directly
proportional to the mass and charge of the particle and inversely
proportional to its velocity.
Energy absorbed in the medium per unit path length of the particle is called its
Linear Energy Transfer (LET). It is usually expressed in keV/m. The concept of LET is
important as biological effects depend on the rate of energy absorption in the medium.
Since alpha particles are doubly charged and are of comparatively heavy mass,
they have a high specific ionization. Hence alpha particles lose energy in matter,
relatively rapidly, by these processes. As alpha rays from a given radionuclide are all
emitted with the same energy, they will have approximately the same range in a given
material. The range of an alpha particle is usually expressed in cm of air.
Beta particles, however, have only 1/7300 of the mass of alpha particle
and have only unit charge. Therefore, beta particles on entry in any material will be
scattered more and thus will have a tortuous path. They penetrate further into the
material. Beta particles have lower specific ionization and LET than alpha
particles. The range of beta particles in any medium is a function of the beta particle
energy and of the density of the medium. An empirical relation between the range (in
8 . 13
g/cm2 of aluminium) and energy (in MeV) of electrons of energies more than 0.6 MeV is
given by the relation.
Rmax(g/cm2) = 0.53 Emax – 0.106
Radiative Collision
When a fast moving charged particle passes close to a nucleus, it undergoes
deflection and loses energy in the form of electromagnetic radiation (Fig.2.2). Radiation
thus emitted is known as bremsstrahlung. Continuous X-rays from an X-ray equipment
are produced by this process. Production of bremsstrahlung is an important consideration
in the shielding of high energy beta particle. The intensity of bremsstrahlung
increases with the atomic number of the medium and decreases with
increase in the mass of the particle. Hence, energy loss by radiation is more
important in heavy elements than for light particles such as electrons.
Beta Particle
hν
Nucleus
Bremsstrahlung
Fig 2.2 : Process of bremsstrahlung
Range of Charged Particles in Matter
After losing all its kinetic energy, a charged particle comes to reset in the
medium. The distance traveled by the particle before coming to rest is known
as its range. The range of a particle depends upon the energy, charge, and mass of the
particle as well as the density and atomic number of the medium. Since alpha particles
lose energy rapidly they can travel only very short distances i.e. their penetrating ability
is very small. An alpha particle of energy 5 MeV has a range of less than 5
cm in air. Even a thin sheet of paper can stop these particles and they cannot penetrate
the human skin. Since beta particles have less mass and lower charge, they lose much less
8 . 14
energy per collision than alpha particles of the same initial kinetic energy and hence they
have a larger range. A beta particle of energy 1 MeV has a range of nearly 40
cm in air and 0.5 cm in tissue.
ELECTROMAGNETIC RADIATIONS
These are wavelike disturbances, which arise in association with vibrating electric
charges. Radio-waves, infrared radiations, visible light, ultraviolet radiation, x-rays and
gamma rays, etc., belong to the class of electromagnetic radiations. They differ only in
their wavelength or frequency. Wave length and frequency are related to each other
by the relation.
C = 
where  is wavelength in cm,  is frequency in hertz and C is the velocity of
electromagnetic radiation (3 x 108m/sec in vacuum). Energy of electromagnetic radiation
is given by the relation,
E = h
where h is a constant called Plank’s constant.
Among electromagnetic radiations only X-rays and gamma rays have sufficient
energy to ionise matter.
Gamma rays and X-rays (also called photons) interact with matter in a variety of
ways, the three main processes of which are: 1) Photoelectric absorption, 2)
Compton scattering, and 3) Pair production.
Photoelectric Absorption (effect)
In photoelectric process all the energy of the incident photon is transferred to an
atomic electron which is ejected from its parent atom. The photon is completely
absorbed. The vacancy created by the ejected electron can be filled by an outer orbital
electron with emission of characteristic X-rays. There is also the possibility of Auger
electron produced by absorption of characteristic X-rays internally by the Atom. The
probability of photoelectric effect decreases with increase of the energy of the photon, but
increases with the atomic number of the medium.
8 . 15
Hence for high atomic number
materials such as lead, at energies lower than 100 keV, the predominant interaction is
photoelectric effect. Figure 2.3(a) shows schematically the photoelectric process.
A photon of energy E will release an electron with kinetic energy Ee = E-,
where  is the binding energy of the electron in that particular orbit.
Some salient features of photoelectric process
1. Photoelectric process involves bound electrons.
2. The probability of ejection of an electron is maximum when the photon energy is
just higher than the binding energy of the electron.
3. The photoelectric absorption coefficient varies with energy, approximately as
1/E3.
4. Photoelectric absorption coefficient varies approximately as the third power of the
atomic number of the absorber (Z3).
5. As the photon energy increases there is greater probability for photoelectron to be
ejected in the forward direction.
Compton scattering
Compton process involves transfer of a part of the energy of the incident
photon to a free electron. The outermost electrons of an atom, which have very low
binding energies, are considered free electrons.
Since Compton scattering
involves these free electrons, the process is independent of the atomic number of
the medium in which the interaction takes place. Since many materials have
approximately equal number of electrons per gram (3 x 1023), absorption by this process
is nearly equal for equal masses of such materials. The photon transfers only a part of its
energy to the electron and gets scattered with reduced energy [Figure 2.3(b)]. The energy
given to the Compton electron is ultimately absorbed in the medium.
Some salient features of Compton interaction
1. This process involves a photon and a free electron.
2. Mass attenuation coefficient for the process is independent of Z of the
medium.
3. In this interaction, probability decreases with increase in energy of the incident
photon.
8 . 16
4. In this interaction, some energy is absorbed and the rest is scattered depending
upon the angle of scattering.
5. In soft tissues, in the energy range of 100 keV to 10 MeV, this interaction is
more predominant
than
the photoelectric or pair production
processes.
6. As the energy of the incident photon is increased, the electron will be ejected
more in the forward direction and it will carry larger portion of the energy.
Pair Production
In the intense electric field close to a nucleus an energetic photon may be
converted into a positron-electron pair. This is known as pair production and is an
example of conversion of energy into matter, which is shown schematically in figure
2.3(c). The minimum photon energy required for this process to occur is 1.02
MeV and the excess photon energy is shared as kinetic energy between the electron and
positron. The positron at the end of its track would encounter an electron. The two
particles annihilate and produce two photons each of energy 0.51 MeV.
Probability of pair production increases with increasing photon energy beyond
the threshold and also with the atomic number of the material. However, at energies less
than 10 MeV it is not an important mode of interaction with most materials.
Some salient features of pair production
1. It is an interaction between a photon and the nucleus.
2. Threshold energy for this process is 1.02 MeV.
3. Probability for this process increases with the square of the atomic number
(Z2) of the medium.
4. This process increases with increase in energy of the photon.
Thus all the three interactions result in the photon energy being transferred to
electrons which subsequently lose energy to the medium as described above.
Attenuation of Gamma Radiation in Matter
When gamma radiation traverses matter it can undergo all the three interactions
with varying probabilities. Part of it may be absorbed, part scattered and part may be
8 . 17
transmitted without undergoing any interaction. This is shown schematically in figure
2.4. The dominating process depends on the energy of the radiation and the nature of the
medium. There is a certain probability for the occurrence of each process and this
probability is referred to as photoelectric, Compton and pair production
attenuation coefficients. The total attenuation coefficient is the sum of these three
coefficients.
Photoelectric Effect
Compton scattering
Compton
Electron
Incident Photon
Incident
Photon
Scattered
Photon
Photoelectron
(b)
(a)
Pair Production
h 0.51 MeV
Incident Photon
(c)
h 0.51 MeV
Fig. 2.3 : Illustration of Photon Interactions
The amount of radiation transmitted through matter decreases with thickness and
can be described by the exponential relation,
I = I0 e-x
8 . 18
Incident photon Fluence
Absorber
Transmitted Photons
Fluence
Scattered Photons
Collimator
Fig. 2.4: Attenuation of Electromagnetic Radiation
Where I0 is the incident intensity of the radiation, I is the transmitted intensity, x
is the thickness of the material and  is called total attenuation coefficient. If x is
expressed in units of cm,  is expressed in 1/cm (cm-1) and is called total linear
attenuation coefficient. The numerical value of  represents the natural logarithm of the
fraction of the incident radiation intensity attenuated by unit thickness of the material.
The quantity /ρ is called mass attenuation coefficient; where ρ is the density of the
medium.
It may be noted that this law is similar to the one describing the decay of
radioactivity with time and an infinite thickness is required for complete attenuation. The
differences in the mode of attenuation of charged particles and that of photons may be
noted. The practical implications of the attenuation law will be discussed in the (chapter 7).
The total linear attenuation coefficient depends upon both the energy of the
photon and the atomic number of the attenuating medium. The variation of total mass
attenuation coefficient with energy for lead and water is shown in figure 2.5.
From figure.2.5 it may be noted that:
1. At low energies, the main interaction is photoelectric interaction.
2. In the medium energy region, the major interaction is Compton process.
3. The pair production starts at 1.02 MeV and increases with energy.
8 . 19
4. There is a dip in the curve, which corresponds to a transition from decreasing
predominance of Compton interaction and increasing predominance of pair
Mass Attenuation Coefficient (cm2/g)
production process.
10
Lead
1.0
0.1
Water
0.1
1.0
10
100
Photon Energy (MeV)
Fig. 2.5: Mass Attenuation Coefficient of water and Lead for different Photon Energy
--------------
8 . 20
3. RADIATION QUANTITIES AND UNITS
A unit is necessary for the measurement of any physical quantity. The
International Commission on Radiation Units and Measurement (ICRU) reviews and
updates, from time to time, the concepts related to quantities and their units in radiation
physics, dosimetry and radiological protection.
Some of the quantities of interest are activity, air kerma, exposure, absorbed
dose, equivalent dose, effective dose, collective equivalent dose, Annual Limit of
Intake (ALI) and Derived Air Concentration (DAC). In 1980, the ICRU
recommended SI units for the above quantities. These new units along with the
corresponding old units are discussed in the following paragraphs.
ACTIVITY, ‘A’
The activity, A of a radioactive material is a measure of its spontaneous
transformation. It is defined as the average number of spontaneous nuclear transformation
(or disintegration) taking place per unit time.
The special name of the unit of activity is Becquerel (Bq)
1 Bq = 1 disintegration per second = 1 dps
The old unit of activity is Curie (Ci)
1 Ci = 3.7 x 1010 disintegration per second
= 3.7 x 1010 Bq
= 37 GBq
Both the indirectly ionising (photons and neutrons) and directly ionising (charged
particles) radiations transfer part or all of their energy when they interact with matter.
KERMA, ‘K’ (Kinetic Energy Released per unit Mass)
The field of indirectly ionising radiation at any point in matter is given by the
quantity Kerma, ‘K’ which is defined as the sum of the initial kinetic energies of all
charged particles liberated by radiation in a material of mass 1 kg.
SI unit of Kerma is Gray and 1 Gy = 1 J Kg-1
When the reference material is air, the quantity is called air kerma.
8 . 21
EXPOSURE, ‘X’
Exposure, ‘X’ is defined as the absolute value of the total charge of the ions of
single sign produced in air when all the secondary electrons (inclusive of positrons and
electrons) liberated by photons in air of mass m are completely stopped in air.
The unit of exposure is C Kg-1 (C/Kg)
With the present technique, it is difficult to measure exposure when photon energies
involved lie above a few MeV or below a few KeV.
The unit of exposure in use is Roentgen, ‘R’, which is defined as that amount of x
or gamma radiation, which would liberate 1 esu of charge of either sign in 1 cc of air
at STP.
1R (Roentgen)
= 1 esu of charge liberated per cc of air at STP
= 2.58 x 10-4 C Kg-1 (air)
Except at very high energies, the exposure defined above is the ionisation
equivalent of the air kerma. And by definition exposure is a quantity restricted to
photons and to air as the medium.
DOSE, ‘D’
The effects (physical, chemical and biological) of radiation depend not only on
the energy transferred to the medium, but also on the energy absorbed by it. The quantity
absorbed dose (or simply dose) is defined as the amount of energy absorbed per unit mass
of the medium at the point of interest.
The SI unit of dose is Gray (Gy) and 1 Gy = 1 J Kg-1
The old unit of dose is rad which is equal to energy absorption of 100 ergs
per gram of material.
1 rad = 100 ergs gm-1
1 rad = 10-2 J Kg-1
1 rad = 10-2 Gy
EQUIVALENT DOSE, ‘HT’
8 . 22
Effects of radiation depend not only on the amount of energy absorbed, but also
on the spatial distribution of ion pairs. Hence, the biological damage caused by the same
dose of different radiations may be different if they have different rates of energy loss per
unit of path length, which in other terms referred as Linear Energy Transfer (LET).
Alpha particles, because of their high energy, charge and mass, cause greater ionisation
per unit path length than gamma radiations, which mediate through singly charged
electrons. One gray of alpha dose is found to be more effective than one gray of gamma
dose. Hence, in radiation protection, to account for this variation in the effectiveness of
different types of radiation, radiation-weighting factor (wR) is used to multiply the
absorbed dose due to each type of radiation. The weighted absorbed dose is called
equivalent dose HT.
i.e. HT = R DT,R wR
where, DT,R is the absorbed dose in tissue for radiation R of radiation weighting factor
wR.
Since wR is a dimensionless quantity, the unit of dose equivalent is also J
Kg-1 (J/Kg). Radiation weighting factor was formerly called quality factor (QF). The
special name for the unit of equivalent dose is Sievert (Sv).
For radiation protection purposes, 20 mGy of gamma dose, 1 mGy of alpha
dose and 2 mGy of fast neutron dose are equivalent.
It should be noted that this is true only for low equivalent dose levels such as
milli-sievert (mSv), centi-sievert (cSv) and applicable only for radiation protection
purposes.
Equivalent dose in Sv = (Dose in Gy) x (wR)
Formerly, the unit of dose equivalent was roentgen equivalent man (rem)
Dose equivalent in rem = (Dose in rad) x (wR)
In radiation protection 1 Sv is too large a quantity. Hence dose equivalents are
expressed in units of mSv (10-3 Sv). On the other hand, Bq is too small a quantity for
many applications. Hence, the amount of radioactivity is expressed in MBq (106 Bq),
GBq (109 Bq) etc.
8 . 23
Table - 3.1 shows wR values for different types of radiations. (The values of wR
reflect the Relative Biological Effectiveness, (RBE, a term used in radiobiology, for
different types and energies of radiation in production of stochastic effects)
EFFECTIVE DOSE, ‘E’
Exposure to radiation may occur to whole body (uniform irradiation) or to
individual organs of the body (non-uniform irradiation). Non-uniform irradiation will
have to be restricted in order to avoid not only deterministic effects but also stochastic
effects. The ICRP recommends dose limits (DL) for stochastic effects and deterministic
effects.
Table - 3.1 Radiation Weighting Factors (wR)
Type and Range
Energies
Photons
all energies
Rad. Weighting
Factor
1
Electrons, muons
all energies
1
energy < 10 keV
5
10 keV – 100 keV
10
100 keV – 2MeV
20
> 2 MeV to 20 MeV
10
Neutrons
Protons, other than recoil protons
Alpha particles, other than fission
> 20 MeV
5
> 2MeV
5
All energies
20
fragments of heavy nuclei
Table - 3.2: Old and New Radiological Units
Quantity
Old Unit
SI Unit
Relationship between units
Radioactivity
Ci (Curie)
Bq (Becquerel)
1 Bq = 2.7 x 10-11 Ci
R(Roentgen)
C Kg-1(Coulombs/ Kg)
Dose
rad
Gy (Gray)
1 Gray = 100 rads
Equiv. Dose
rem
Sv (Sievert)
1 Sv = 100 rems
Exposure
8 . 24
1 R = 2.58 X 10-4 C/Kg
Effective Dose
rem
Sv (Sievert)
1 Sv = 100 rems
If several tissues, T1, T2, T3, etc., individually receive equivalent dose HT1, HT2,
HT3, etc. then the total risk to the individual should not exceed that resulting from the
stipulated dose limit to uniform whole body irradiation. A number of organs are
considered on the basis or their sensitivity and the seriousness of the damage. Risk
factors are age and sex dependent. Depending on the extent to which the risk from
stochastic effects in a tissue/organ may contribute to the total risk from stochastic effects,
a weighting factor, called the tissue-weighting factor, wT is assigned to each
tissue/organ. Thus the effective dose (E) is defined as,
E = T HT . wT Sieverts
wT represents the contribution of tissue T to the total risk due to stochastic effects
resulting from uniform irradiation of whole body. (Table 3.3)
ANNUAL LIMIT ON INTAKE, (ALI)
ALI means the greatest value of the annual intake of the specified radionuclide
that would result in a committed dose equivalent not exceeding the annual dose
equivalent limit, prescribed by the Competent Authority, even if intake occurred every
year for 50 years. ALI values are given for ingestion if intake is through mouth and for
inhalation if intake is through inhalation. ALI values for some of the important radio
nuclides used in research are listed in table 1.1.
DERIVED AIR CONCENTRATION, (DAC)
DAC means the maximum concentration of a radionuclide in the ambient air
which, if inhaled by a person for 2000 hrs in a year, at a breathing rate of 1.2 m3/h, will
not result in annual effective dose equivalent in excess of the limits prescribed by the
competent authority.
DAC = ALI / 2.4 x 103
Bq/m3
Table-3.3 Tissue Weighting Factors
8 . 25
Tissue /Organ
Tissue Weighting Factor, (wT)
Gonads
Bone marrow
Colon
Lung
Stomach
Bladder
Breast
Liver
Oesophagus
Thyroid
Skin
Bone surface
Remainder*
* Other tissue/organ which are not included in the table
------------
8 . 26
0.20
0.12
0.12
0.12
0.12
0.05
0.05
0.05
0.05
0.05
0.01
0.01
0.05
4. RADIATION DETECTORS & MONITORS
RADIATION PROTECTION INSTRUMENTS
Ionising radiations cannot be seen, felt or sensed by human body in any way but
excessive exposure to them may have adverse health effects. In order to avoid excessive
exposure, appropriate and efficient radiation-measuring instruments are needed. It is not
only important to measure (monitor) the radiation exposure at a place where there is a
potential of radiation exposure but also the instrument used for monitoring must be
appropriate and easy to interpret the results with high degree of accuracy.
Radiation measuring instruments are needed to detect and quantify two types of
potential exposure: external exposure due to penetrating radiation emitted by the
radioactive sources lying nearby at the workplaces; and internal exposure from the
radioactive material that has got entry into the human body inadvertently while working
with it.
There are four basic types of radiation measuring (monitoring) instruments that
are used in a research lab handling radio-nuclides:
a. Dose rate meters: used to measure the potential external exposure rate
b. Dosimeters: used to measure cumulative external exposure
c. Surface contamination meters: used to measure level of radioactive
contamination on surface to evaluate potential internal exposure when
radioactive substance is distributed over a work surface
d. Airborne contamination meters: used to measure level of radioactive
contamination in air to evaluate potential internal exposure when a
radioactive substance is distributed within an atmosphere
All radiation monitoring instruments consist of following key components:
(a) Detector: The detector contains a medium, which absorbs radiation energy
and converts it into a signal. The signal can be electric charge, light,
chemical change etc. The medium generally used for radiation detection
are –
i.
Gases (Ionisation Chambers, Proportional Counters, GeigerMuller Counters)
8 . 27
ii.
Scintillation Media [NaI(Tl), Anthracene etc.]
iii.
Photographic Emulsions [Film]
iv.
Solid State Detectors [Semiconductors,
Thermoluminescent Phosphors]
(b) Amplifier: The signals from a detector need to be electronically
amplified
(c) Processor: According to the type of instrument, the processor is a
device to measure the size or number of signals produced in the detector.
It may also translate the quantity measured into appropriate radiological
units
(d) Display: The measurement is presented either in a digital format or an
analog display by a pointer on a graduated scale.
GAS FILLED DETECTORS
This is a most common type detector used in radiation monitors. These detectors
are filled with gas and normally are cylindrical in shape. They have two electrodes, the
central and the outer sheath, separated by an insulator. Since radiation produce
ionisation in a gas, on exposure to radiation, positive ions and electrons are
produced inside the detector volume. A variable positive voltage is applied to the central
electrode with respect to outer sheath. The positive and negative ions drift to the
negative and positive electrodes respectively under the influence of electric field.
The ionisation current is measured in outer circuit, which is proportional to the
number of ion pairs produced per second. Depending on the strength of the
electric field the detectors are classified into ionisation chamber,
proportional counter, and Geiger-Mueller (GM) counter.
(a) Ionisation Chamber
It is a versatile device. The type of gas filled in chamber, the pressure of
gas and size of the chamber depends on the purpose of use and radiation
intensity to be measured. The ionisation chambers used for radiation monitoring are
mostly filled with air at atmospheric pressure. These detectors are mostly used for
8 . 28
measuring high radiation level normally in the range of 10 - 50 mSv/h. Small
ionisation chambers are often used for personnel monitoring such as pocket dosimeter.
(b) Proportional Counter
It is also a gas filled detector. In this detector (counter), applied voltage across the
electrodes is higher than that applied in ionization counter. In this counter, the initial
ionisation produced in the gas due to radiation causes further (secondary) ionisation
Radiation
(+ve) Central Electrode
Display
1.56 mGy/h
Insulator
(-ve)Outer Electrode
Electron
High
Voltage
Positive ion
because of the existence of higher electric field across the electrodes. Therefore, larger
number of ion pairs are produced and make the counter more sensitive than ionisation
chamber (counter). In addition to its use as a dose-rate meter it can also be used for
energy discriminations.
(c)
Geiger-Mueller (GM) Counters
It is also a gas filled and is used most widely in radiation protection. The major
advantages of this detector are : available in wide variety of shapes, large signal output
and insensitive to environmental conditions. The End Window type GM counters
are used for surface contamination measurement in research laboratories. The
counter display in case of GM survey meter gives the exposure-rate, whereas the display
in case of contamination monitor gives the count-rate. These instruments are used
8 . 29
for measuring low radiation level in the range of 2 - 200 µSv/h. It is very useful for
radiation survey of the laboratories. These are not useful in high radiation field or to
measure pulsed radiation.
SCINTILLATION DETECTORS
The gas filled detectors register the ionisation produced by radiation in gas. The
scintillation detectors work on a different principle. In scintillation detector
when radiation deposits its energy in a scintillator, it produces
fluorescence (light flash). The light flashes (fluorescence) produced are used for
measuring radiation parameters such as energy and intensity of radiation. Some
times the scintillators are also called as phosphor.
Scintillation detectors are usually in solid form. But in some applications the
scintillators are used in liquid form also. NaI(Tl), CsI(Tl), ZnS(Ag), Anthracene
and Plastic scintillators are some of the examples of solid scintillation detectors.
NaI(Tl) and CsI(Tl) detectors are used for detecting gamma radiation,
whereas ZnS(Ag) is used to detect alpha radiation and the Anthracene and
plastic scintillators for beta radiation.
The light flashes (scintillation) produced are converted in electric pulses (using
Photo Multiplier Tubes) and then fed into suitable electronic circuitry where the
scintillations are discriminated to analyse different types of radiation and even the
different energies of the radiation.
LIQUID SCINTILLATION DETECTOR (counter)
Liquid scintillation counting (LSC) is widely used for detecting non-penetrating
radiation such as beta radiation of weak (low) energy, alpha radiation and penetrating
radiation such as gamma radiation of very low energy (energy <20 keV). For
this type of radiation, the detection efficiency of LSC is very high as
compared to other detectors. The radioactive samples of radio nuclides emitting such
radiation are mixed with LS (in the liquid scintillating medium). The energy of weak
radiation is absorbed in the scintillating medium, resulting in the molecules to become
excited. The excited molecules emit photons and then return to their ground state. The
8 . 30
light output (intensity of light flashes produced) is proportional to
the radiation energy. The liquid scintillator solution cannot be reused. It is used
only once and treated as radioactive waste (chapter -10) and stored for safe disposal. LS
is most efficient detector for detecting tritium and 14C labelled samples.
LS are a cocktail (mixture) of two organic compounds (scintillators). One is
called the primary and the other secondary scintillator. The function of secondary
scintillator is to shift the wave length of light output from primary scintillator to the wave
length sensitive to photo cathode of photo multiplier tube (PMT) used in the system. PPO
(2-5 di-phenyle-oxazole) with tolune or dioxane as the solvent and POPOP (triphenyl di-oxazole) are one of the examples of primary and secondary scintillators.
Radiation interacts with the molecules of scintillators in the manners described in
chapter-2.
The organic scintillators used in LSC have very high absorption coefficient for
low energy beta, alpha and gamma radiation of energy <20 keV, but very low for
energetic (>20 keV) gamma radiation (refer chapter-2). If energy of gamma radiation is
>300 keV and mass number (Z) of interacting medium is low, as is the case with LS, the
probability of photoelectric effect is very small. Thus the probability of absorption of
gamma radiation in LS by Compton scattering process is very high. As a result the
detection efficiency of energetic gamma radiation will be very poor. For gamma radiation
of energy <20 keV, the energy absorption by photoelectric effect process in LS would be
very high, resulting in high detection efficiency. The counting efficiency for gamma
radiation emitted by 125I can be as high as 76% in a typical emulsifier type LSC.
SEMICONDUCTOR DETECTORS
The solid state detectors, such as silicon and germanium detectors, are mainly
used for gamma spectroscopy (energy spectrum or discrimination). These detectors have
good energy resolution (4 eV). This is because the average energy required to produce
an electron-hole pair in semiconductors is about 3.5 eV as against 35 eV energy to create
ion pair in air. As a result larger number of electron hole pairs would be produced in
semiconductor as compared to number of ion pairs produced in air for same gamma
radiation energy.
8 . 31
Where, the energy resolution of gamma radiation is high in (germanium)
semiconductors, the detection efficiency is low. These detectors are mainly used to
identify gamma emitting radio nuclides in unknown samples by resolving radiation
energy emitted.
THERMO-LUMINESCENT DOSIMETERS (TLD)
When certain solids are exposed to ionising radiations, the electrons released in
the ionisation process are trapped in the lattice imperfections in crystalline solids. These
electrons remain trapped till they are released by thermal agitation at some elevated
temperature. They emit light in the process and the quantity of light emitted as the
material is heated may be measured and related to the absorbed dose in the material. The
material thus exposed and heated, can be reused after proper annealing process.
Many thermo-luminescent materials like LiF3, Al2O3, CaF2, CaSO4:Dy have been
studied. Of these CaSO4 : Dy has been found to be very useful for dosimetric purposes,
due to its high sensitivity, low fading. indigenous production and many other useful
characteristics.
Since the effective atomic number of Li F is comparable to that of tissue and air,
it is found to be almost energy independent. This phosphor is also widely used in both
personnel monitoring as well as in other dosimetric applications. Other TLDs such as Ca
S04 : Dy, Al2 O3, have higher effective atomic numbers and hence show significant
energy dependence to X and gamma rays. Hence metal filters have to be used to
compensate the dependence of response on photon energy. The use of filters also permits
the estimation of energy of photons below 200 keV.
CHEMICAL DETECTORS
Photographic Emulsions Films
Photographic film consists of a sensitive layer of silver halide crystals on gelatine
spread on cellulose acetate base. The thickness of emulsion layer ranges from 10 - 25
microns. When ionising radiation or visible light falls on it a latent image is formed on it.
The radiation exposure causes ionisation in the silver bromide crystals (grains) and a
group of silver clumps containing several silver atoms are formed on the surface of the
8 . 32
film. After developing the film, each exposed grain is reduced to metallic silver. The
developer serves merely as reducing agent. The unaffected, undeveloped silver halide
crystals are dissolved by immersing the film in fixer solution. The processed film shows
blank and blackening on the surface of the film. The amount of blackening is related to
the quantity of radiation recorded by it. The blackening is measured in terms of optical
density. It is related to the quantity of radiation absorbed in the film. The optical density
is measured using an instrument known as densitometer.
This detector is used to record radiation dose (cumulated exposure) of personnel
working with radiation i.e. in personnel monitoring device. These are special
photographic films and are loaded in a special cassette having combination of certain
metallic filters.
RADIATION MONITORING INSTRUMENTS
Radiation monitoring instruments can be broadly classified into three categories
based on their applications as i. area monitoring instruments ii. portable
survey instruments and iii. personnel monitoring instruments. Since
there are various situations i.e, type of radiation present, energy of radiation and level of
radiation to be monitored, it often becomes necessary to use multiple instruments to
monitor the radiological parameters. Almost all the types of detectors mentioned in the
previous sections are used in radiation monitoring.
For accurate measurement of radiation, special ion chambers having uniform
response to incident radiation of all energies are used. Instruments based on such ion
chambers have accuracy better than ± 5% and are normally referred to as
dosimeters. The GM counter and some scintillators detectors have different responses for
different energies of incident radiation. Such instruments have accuracies of the order of
±10% or more. Therefore these are mainly used for carrying out general area survey and
measuring radioactive contamination. Apart from this, TL phosphors and photographic
films are used as personnel dosimeters. These are called passive dosimeters since they
require no power supply for operation and are of integrating type i.e. they keep on
accumulating the radiation dose until readout is made. The instruments based on ion
chambers are normally expensive and delicate. As the ionisation currents are very low
8 . 33
(10-12 to 10-9 A) for a range of exposures from a few mR/h to several R/h, special low
current electrometer amplifiers are used. They require regular maintenance. The G.M.
counter based instruments are rugged, comparatively inexpensive and require little
maintenance. The instruments based on scintillators are expensive and delicate as they
contain the photomultiplier tube and are normally used for measurement of low radiation
levels of the order of μ-rads to m-rads (μSv). The passive dosimeters like TLD and the
photographic film dosimeters are being used for regular personnel monitoring.
A practical instrument must tell us which type of radiation it is measuring, as well
as the intensity. An instrument may indicate that there exists radiation field in a work
location. But this is not enough. It must help to enable us to assess the hazard. For
example, if the radiation being measured is gamma radiation, it would be exposing the
whole body. If the radiation is a beta radiation it would be exposing the external skin. In
earlier situation it would give whole body equivalent dose-rate whereas in the later one it
would be shallow dose rate and not a whole body equivalent dose-rate. The knowledge of
radiation would allow using appropriate radiation shield to eliminate hazard. If the
radiation is pure alpha radiation, there is no external hazard but one have to take
precautions to avoid ingesting alpha emitting radioactivity. Generally, in mixed radiation
fields, numbers of different instruments are used to measure the intensity of each
radiation present.
CALIBRATION CHECK OF RADIATION MONITORING EQUIPMENT
Calibration refers to determination and adjustment of instrument in a particular
radiation field of known intensity. The calibration of survey meters may be done with the
help of a standard source of reasonably high activity (137Cs in MBq strength). The survey
meter is positioned at a known distance from a standard source having no scattering
object nearby and the exposure rate is recorded. The expected value of exposure rate is
calculated using specific gamma ray constant for that radionuclide (Appendix -4 table 1).
The ratio of the expected value to measured value gives calibration factor. Routine
calibration commonly involves one or more sources of a specific radiation type and of
sufficient activity to provide calibration on all ranges of concern. Before calibration the
instrument is checked for radioactive contamination, condition of batteries, loose or
8 . 34
broken parts, proper operation of switches and zero adjustment. Radiation Standards
Section (RSS), Radiation Safety System Division (RSSD), Bhabha Atomic Research
Centre (BARC), in our country, carries out calibration of survey meters. It is desirable
that all the survey instruments are calibrated periodically (at least once in three years or
when ever the unit has undergone major repairs).
In the interval between calibrations, however, the instrument user should validate
acceptable operation by carrying out a performance check. This is merely intended to
establish whether the instrument is functioning within specified limit (i.e. within 20% of
the expected value). This is done by measuring the response of the instrument to a
radioactive source of known activity (137Cs or 60Co of the order of a few kBq activities)
under a specified geometry. The initial performance check should be carried out
immediately after calibration of the instrument. During subsequent performance checks
the same source and same geometrical condition should be employed so as to correlate
the earlier findings. The user of survey instrument should have adequate knowledge of
preventive maintenance of the instrument, which includes the removal of batteries and
keeping inside dehumidified enclosures, when not in use.
-----------
8 . 35
5. BIOLOGICAL EFFECTS OF IONIZING RADIATION
INTRODUCTION
The interaction of ionizing radiation with human body, arising either from external
radioactive sources (outside the body) or internal contamination, leads to biological
effects which may later show up as clinical symptoms.
Both radiobiological
investigations and human exposure to ionizing radiation have contributed to our present
knowledge in this area. Radiobiological data have been derived mostly from microorganism, cultured mammalian cells and whole animal systems. Human data have been
derived from the follow-up of the (a) survivors of atomic bomb explosions in
Hiroshima and Nagasaki, (b) inhabitants of Marshall islands, who were
exposed to fallout from thermonuclear devices, (c) uranium miners, (d)
radium dial painters, (e) pioneer X-ray technicians and radiologists, (f)
patients exposed to radiation for medical reasons; and (g) victims of
nuclear accidents. A careful analysis of these data has yielded reasonable quantitative
estimate of biological effects of radiation in man. Some of these are briefly described in
this chapter.
THE CELL
Cell is the basic unit of all living organisms. A living organism may be made up
of a single cell or many cells; e.g. the organisms which cause diseases like typhoid and
tuberculosis are single celled. Man is a multi-cellular organism having about 1014 cells.
The structure of a typical animal cell is shown in figure 5.1. It has an outer envelope
called cell membrane or plasma membrane. Inside the cell there is nucleus in the
centre. Outside the nucleus is a viscous liquid called cytoplasm. The cytoplasm
contains structures like mitochondria, golgi bodies, lysosomes, etc., which perform
important cellular functions.
The nucleus contains chromosomes which are tiny thread-like structures made
up of deoxyribonucleic acid (DNA) and protein. DNA molecules contain in-coded
form (as sequence of bases), all the information required for the cellular function and thus
control the growth, development and well-being of the individual. Sections of
8 . 36
chromosomes which contain information for specific functions are called genes.
The size and shape of the cells are different in different parts of the body,
depending upon their function. Cells of similar nature are organized together to form a
tissue. Different types of tissues join to form an organ. Different organs form a system;
e.g. respiratory system consisting of nose, wind pipe and lungs; digestive system
consisting of mouth, food pipe, stomach and intestine. The cell can be compared to a
factory. The nucleus is .the control room. So, if the control room is damaged, work in the'
cell factory' slows down or stops.
Cel1s can be grouped into two categories: (1) somatic cells and (2) germ
cells. Somatic cells constitute various tissues such as brain, kidney, skin, liver etc. Germ
cells, also called reproductive cells, are those which participate in the reproductive
process. They are sperms in males and ova in females. All somatic cells in human
body contain 46 chromosomes which occur in 23 pairs. The germ cells contain
only 23 chromosomes - single copy of each pair. Figure 5.2 shows the human
chromosomes (male).
EFFECT OF RADIATION ON CELLS
The basic difference between ionizing radiations and the more commonly
encountered radiations such as light is that the former have sufficient energy to cause
ionization in matter. When it is incident on the body, a part or whole of the energy may
be absorbed by the cell through the process of ionisation and excitation (Refer
Chapter-2). Since the water content in human tissues is more than 75% most of the
energy will be initially deposited in the water molecules and only a small part is taken up
directly by the other bio-molecules. The excited and ionized water molecules undergo a
series of reactions. Radiation on interactions with water produces the radiolytic
products of water such is H, OH, HO2, eraq, O2 and H2O2. These free radicals react
readily with bio-molecules in the cell and result in damage to important biomolecules such as DNA and proteins. Such damages may lead to (a) inhibition of cell
division, (b) chromosome aberrations,
(c) genes mutation, and (d) cell
death. While the absorption of radiation energy in the organ
8 . 37
tissues takes a very short time (only 10-15
sec), the appearance of biological effects
(damage due to absorption of radiation)
may take a few hours to several years.
a. Inhibition of cell division:
Cells
originate and multiply by the process of
cell division. It is one of the basic functions
in all living organisms. Even in an adult
human body, cells in certain organs are in a
constant process of division and thus
Fig. 5.1 : A Typical Cell.
impair the function of tissue and organ.
Fig. 5.2 Human chromosomes (Male 44 + XY)
Fig. 5.3: Human chromosome (Lymphocyte) with aberrations
(d: dicentric, Ace: acentric fragments, r : ring)
8 . 38
b. Chromosome aberrations: Radiation can cause breaks in chromosomes. Majority of
the breaks may get repaired and the damage may not manifest. However, certain
breaks may lead to rearrangements of genetic material which can be seen under a
microscope. Such events are called chromosomal aberrations (Figure 5.3), The frequency
of various types of di-centric chromosome aberrations can be correlated to dose and
hence can serve as a biological dosimeter.
c. Gene mutation: Alterations in the information content of genes (DNA) are known as
gene mutation. Damage to chromosomes (Chromosome aberrations) may lead to
change the information content of DNA.
d. Cell death: Irradiation can lead to cell death as a result of any or all of the above
effects. Cell death is usually expressed as fraction of cells surviving after a given
exposure. The effect of dose on mammalian cell survival is shown in figure 5.4(a). At
low doses the survival curve has a broad shoulder where cell survival decreases slowly
with dose. This is attributed to the ability of cells repair radiation damage. However, at
higher doses, where repair capacity is saturated, the survival decreases rapidly with
increasing dose, in an exponential manner. Figure 5.4(b) shows the dose response of
human peripheral blood lymphocytes for the induction of chromosomal aberrations.
Depending upon the type of radiation it may increase linearly or nonlinearly with
increasing dose. A typical metaphase spread of human lymphocyte carrying
chromosomal aberration is shown in figure 5.3.
Fig. 5.4: (a) Effect of dose on cell survival. (b) Calibration curve for induction of
dicentric
8 . 39
BIOLOGICAL EFFECTS OF RADIATION IN MAN
The detrimental effects of radiation in human body are produced as a result of
damage to the individual cells. These may be divided into two classes, namely somatic
effects and hereditary effects. The somatic effects arise from damage to cells in a
particular irradiated tissue and affect the irradiated person only. The hereditary effects are
due to damage to germ cells which may manifest in the progeny of the irradiated
person.
Somatic Effects (Early effects): Somatic effects of radiation may appear immediately
after exposure (within a few hours to weeks) or much later (years or decades after
exposure). The early effects are due to an acute exposure, i.e., large doses received over a
short period of time (a. few hours or less) and attributed to depletion of cell population
due to cell-killing. Acute exposure of whole body to about 1 Gy may lead to reduction
in lymphocyte and granulocyte counts and radiation sickness in the form of
nausea and vomiting. These are however transient and the exposed person recover
after one or two days. But the severity of the effect increases with radiation dose. At
doses higher than 3 Gy, irrecoverable damage occurs to the blood forming
organs (bone marrow, spleen, lymph node etc., with the possibility of death in a
few weeks time. In the dose range of 3-5 Gy about 50% of the exposed persons
may die within 60 days (LD50/60). Anemia, infection, and high fever are the main
symptoms. These symptoms are called Haematopoietic Syndrome.
At higher doses, in the range of 7 to 10 Gy, cells in the gastrointestinal
system get severely damaged leading to diarrhoea, loss of appetite, dehydration,
electrolyte imbalance, weight loss and high fever. These symptoms are typical of
Gastrointestinal Syndrome (GIS). Death occurs in 7 to 14 days time.
As the dose is increased further, consequences of damage to central nervous
system manifest. At doses in the range of 25 Gy and above, the damage becomes
so severe that depression, fatigue, delirium and comma appear and death
occurs within a few hours to 2 days. These symptoms are known as Central Nervous
System Syndrome (CNS).
The quickness of occurring early symptoms such as nausea, vomiting etc and the
8 . 40
degree of severity is good indications of the level of exposure. Persons exposed to doses
in the range of LD50/60 exhibit these symptoms within one or two hours.
The nature and seriousness of the early effects also depend upon whether the
exposure is to the whole body or part of the body. Even though whole body irradiation to
3-5 Gy may kill 50% of the exposed persons, the same dose given to a part of the
body will cause only local effects. Some of the local effects are reddening of the
skin (erythema), hair falling off (epilation), temporary or permanent sterility (when
reproductive organs are exposed). Table 5.1 lists some of the early effects of whole body
and local acute exposures.
All the early effects exhibit a threshold dose below which the effects do
not occur. Beyond the threshold dose, severity of the effect increases with increase of
dose. Hence, these are referred to as deterministic effects. Because of the large
threshold doses, acute effects do not occur from exposures arising from
normal work with radiation. However, the lethal effects of radiation have been
exploited in a number of beneficial applications such as in industry and medical where
the radioactive sources in giga or tera Bequeral quantities are used for example:
sterilization of medical products, food preservation, treatment of cancer, and treatment of
effluent from municipal sewage.
Late effects: Persons who recover from early effects may still develop some other types
of effects later in life. Similarly exposure to low levels of radiation over long periods of
time, which cannot produce any early effects, may also lead to late effects. Late effects
are characterized by a long latent period, which can be as long as 30-40 years. The
important late effects are cancer, fibrosis in various tissues and cataract of the eye lens.
Cataract is progressive loss of vision. The transparent cells in the eye lens,
which facilitate vision, are killed by high doses of radiation. Since there is no blood flow
through lens, the dead cells are not removed. Accumulation of dead cells results in the
opacity of lens, which lead to the impairment of vision.
8 . 41
Table 5.1 : Biological Effects of Acute Exposure to Radiation
A. Whole Body Irradiation (low LET radiation)
Dose Range
Less than 0.1 Gy
Above 0.1 Gy
Above 0.5 Gy
Above 1 Gy
3-5 Gy
5-10 Gy
Immediate Effect
No detectable effect
Chromosome aberrations detectable
Above effect plus transient reduction in WBC
count granuclocyte count.
All the above plus nausea, vomiting, diarrhea
(NVD), loss of appetite, radiation sickness:
recovery probable.
All the above with increased severity. Death of 50
percent of exposed population in about 60 days.
Severity of above effects increases. Almost 100
percent death (at the higher dose).
B. Local Irradiation (Low LET radiation)
Dose
0.15 Gy
3.5-6.0 Gy
1.5-2.0 Gy
2.5-6.0 Gy
3 Gy
5 Gy
6 Gy
10-20 Gy
Region
Testes
Testes
Ovaries
Ovaries
Hair follicles
Eye
Skin
Skin
Effect
Temporary sterility
Permanent sterility
Temporary sterility
Permanent sterility
Epilation (temporary)
Cataract (after 5-10 years)
Skin erythema/depilation
Burns, blisters, wounds, death of
tissue (necrosis), Permanent loss of
hair.
Cancer results from viable cells which have received damage to their control
system in the form of gene mutations or chromosomal aberrations. Such damage may
lead to over proliferation of cells in an organ resulting in cancer. Cancer cells multiply
and grow in an uncontrolled manner. Animal experiments and human
epidemiological studies have shown that radiation, at high doses, is carcinogenic.
Cancer mortality data based on the follow-up of atom bomb survivors of Japan are
presented in tables 5.2 and 5.3. At high doses (> 0.2 Sv) there is a clear indication that the
leukemia incidence increases with bone marrow dose. However, the shape of the
dose-effect relationship at low dose levels is not well established. Because of
8 . 42
the high frequency of naturally occurring cancers, it is not possible to demonstrate
conclusively whether or not small doses of radiation are carcinogenic to man.
Epidemiological studies in high back-ground radiation. Areas around the world do not
reveal any increased risk of cancer to the population.
Similarly, among the survivors of atomic bomb explosion in Hiroshima and
Nagasaki, no statistically significant increase in leukemia incidence rate has been
reported in population groups exposed to doses below 0.2 Sv. However, to be on the
conservative side in radiation protection, the probability of radiation induced cancer is
assumed to increase proportionately with dose, without any threshold. Accordingly, any
dose, however small is taken to be associated with a certain level of risk.
This kind of effect is referred to as Stochastic which means ‘random or statistical
nature'.
A number of human population groups exposed to high levels of radiation have
been studied to estimate the risk of radiation induced cancer at low levels of exposure.
The current estimate, suggested by the International Commission on Radiological
Protection, is that there would be 50 excess cases of cancer per million populations if
exposed to 1 mSv of radiation dose. This implies that if one million persons (both men
and women) each receive a dose of 1 mSv, the number of fatal cancers attributable to
radiation would be 50 and these appear over a period of a few decades. To put these
estimates into perspective, it should be noted that in a typical population of 106 people,
there are 1 - 2 x 105 deaths from spontaneous cancers.
.
Hereditary Effects: Hereditary effects (another example of stochastic effect), occur in the
progeny of exposed persons, if the germ cells carrying radiation induced damages
(mutations, chromosomal aberrations, etc.) participate in the process of fertilization. Only
that exposure of the reproductive organs which occurs up to the time of conception can
affect the genetic characteristics of the off-spring. There is no human data which
demonstrate that radiation induces hereditary defects in man. However, based on animal
experiments, the ICRP has estimated the risk of serious genetic disorders in future
generations following irradiation of either parent, to be about 10 per million live births
per milli-sievert (mSv). To put these estimates into perspective, it may be noted that more
than 500 types of human diseases are attributable to genetic factors and nearly 10
8 . 43
percent of all new born children suffer from spontaneous genetic disorders. It may be
mentioned that mutations induced by various agents like heat and chemicals
etc. are indistinguishable from that induced by radiations. Table 5.4 gives a summary of
the life-time risk of radiation induced stochastic effects in man.
Table 5.2 : Observed and Expected Deaths for solid Cancers in Japanese A-bomb
survivors
Colon
Subjects
Observed
Dose (Sv)
Deaths
0(<0.005)
36,459
3,013
0.005 - 0.1
32,849
2,795
0.1 - 0.2
5,467
504
0.2 – 0.5
6,308
632
0.5 – 1.0
3,202
336
1.0 – 2.0
1,608
215
> 2.0
679
83
Total
86,572
7,578
Donald Pierce et.al (RERF update, vol 8, p.10)
1950-1990
Expected
Background
3,055
2,710
486
555
263
131
44
7,244
Excess
Deaths
-42
85
18
77
73
84
39
334
Table 5.3 : Observed and Expected Deaths of Leukemia
Marrow
Subjects
Observed
Dose (Sv)
Deaths
0(0.005)
35,458
73
0.005 – 0.1
32, 915
59
0.1 – 0.2
5,613
11
0.2 – 0.5
6,342
27
0.5 – 1.0
3,425
23
1.0 – 2.0
1,914
26
> 2.0
905
30
Total
86,572
249
Donalt Pierce et.al (RERF update, vol.8, p.10)
1950-1990
Expected
Background
64
62
11
12
7
4
2
162
Excess
Deaths
9
-3
0
15
16
22
28
87
Table 5.4 : Estimated Lifetime Risk of Radiation Induced Stochastic Effects in Man
(Chronic Exposure)
Effect
Risk per million per mSv
Cancer
Hereditary
50
10
----------
8 . 44
Spontaneous risk per
million
1 x 105 – 2 x 105
10% of all live born
children
6. OPERATIONAL LIMITS
INTRODUCTION
All human endeavors entail risk of some type or the other. In the same manner
ionizing radiation too pose a physical risk to people, who may be exposed by natural and
artificial means. The interaction of radiation with matter produces ionisation phenomena
capable of modifying the chemical behavior of molecules. If this occurs in live cells,
biological effects of varying degree of severity may result in. People who work with
sources of ionizing radiation, some members of the public and patients who undergo
radiological procedures are exposed to risk of ionizing radiation. It is not possible to
eliminate these risks totally, but it is feasible to control them within
acceptable limits by following the principles of radiation protection. Hence, it is
necessary to develop safety standards to work with radiation so that the risks are kept
minimum.
The International Congress of Radiology established the International
Commission on Radiological Protection (ICRP) in 1928. ICRP issued its recent
basic recommendations on Radiation Protection in 1991- 'ICRP Publication
No.60'. These recommendations led to the joint publication of the International Basic
Safety Standards for Protection against Ionizing Radiation and for the
Safety of Radiation Sources (BSS). The objective of protection from ionizing
radiation is to prevent from radiation induced deterministic effects and to reduce the
likelihood of stochastic effects by considering economic and social factors. There
are three basic principles that sum up the philosophy of radiation protection and these are
discussed below.
PRINCIPLES OF RADIOLOGICAL PROTECTION
Based on the above consideration, the principles of radiological protection are
Justification of Practice, Optimisation of Practice and Dose
Limitations
8 . 45
JUSTIFICATION OF PRACTICE
This is the first principle, which emphasises on justification of applications of
radiation in a practice. It is judged from the total detriment from a proposed
practice involving exposure from ionising radiation should be less than the
expected benefit. On that basis the application of radiation in medicine, industry,
agriculture and research can be justified but the application of radioisotopes in
toys and jewellery cannot be justified.
OPTIMIZATION OF PROTECTION
Implementation of protection measures calls for value judgment. The greater the
level of protection, the higher the degree of safety achieved. At the same time, greater
levels of protection would involve expenditure, which may reduce the ultimate value of
practice. For example research laboratories carrying out in-vitro experiments with small
amount of radioactivity and week energetic radiation don't call for additional structural
shielding for assuring prescribed degree of safety. But the laboratories, where large
activities are handled/stored need adequate structural and local radiation shields.
Providing additional structural shielding against radiation, a greater degree safety may be
achieved but this measure would achieve a small increase in degree of safety for a large
increase in the expenditure. Hence, it is recommended that protective measures should be
optimized that the radiation exposures are kept As Low As Reasonably Achievable
(ALARA). This appropriate socio-economic consideration should be factored in
optimizing the level of protection.
DOSE LIMITATION
In order to reduce the magnitude of the risks associates with a justified practice,
limits on individual doses are established to prevent the occurrence of deterministic
effects and minimise the likelihood of stochastic effects. The
monitoring of these limits should take into account of doses generated by external
sources and those produced by the intake of the radio nuclides into body. The dose limits
applicable to workers and to members of the public are given in table 6.1 and discussed.
It must be noted that the exposures due to natural radiation background and
medical procedures are excluded from the dose limits.
8 . 46
DOSE LIMITS FOR RADIATION WORKERS
In setting the dose limits for radiation workers, ICRP has ensured that, for a
continued exposure at that level, the estimated risk is not unacceptable. By considering
the total harm arising from somatic (fatal cancer, non-fatal cancer and hereditary effects),
the commission recommends a limit to the effective dose of 20 mSv per year
averaged over 5 years (100 mSv in 5 years), with further provision that the
effective dose should not exceed 50 mSv in any single year. The 5-year
period refers to a discrete 5 year calendar period. It is implicit that the
dose constraint for optimisation (e.g. planning of installation) should not exceed 20 mSv
in a year. The effective dose limit ensures the avoidance of deterministic effect in all
body tissues and organs (since the equivalent doses in all cases are less than the threshold
for any deterministic effects) except the lens of the eye, which makes a negligible
contribution to the effective dose and the skin which may be subjected to localised
exposures. Separate equivalent dose limits are needed for these tissues. The annual limits
are 150 mSv for the lens and 500 mSv for the skin, averaged over any 1 cm 2 regardless of
the area exposed.
INTERNAL EXPOSURE
A person may receive radiation doses from external and/or internal sources. In
research application, where radio nuclides in unsealed forms are used, radio nuclides may
enter the body through injection, inhalation, wounds, or absorption by skin.
The dose limit is the same, irrespective of the exposure route-internal, external
or combination. Accumulation of internally deposited radio nuclides will depend on
radioactive contamination present in air and water, the rate of intake and metabolism.
Limits are therefore prescribed in terms of Annual Limits on Intake (ALI) for
different radio nuclides if the exposure is only due to intake of activity. ALI is based on a
committed effective dose of 20 mSv. Dose limits are only the part of a system of
protection aimed at achieving levels of dose that are as low as reasonably
achievable, economic and social factors being taken into account. It is not to be seen as
a target because; if received over a lifetime may lead to risk level verging on becoming
unacceptable. However, any isolated exposure exceeding the limit need not cause alarm,
8 . 47
but a thorough examination of the design and operational aspects of protection must be
conducted. If the dose is unknown, or is thought to be high, referral to a physician should
be considered.
If it is likely to receive radiation exposure from external source and from the
internally deposited radioactivity, both the exposures should satisfy the following
equation.
Intake
E.E

1
ALI
ADL
Where: Intake - means through both ingestion and inhalation
ALI – ALI both ingestion and inhalation
E.E –
External exposure
ADL – Annual dose limit
OCCUPATIONAL EXPOSURE OF WOMEN
The basis for the control of occupational exposure of women who are not
pregnant is the same as that for men and the commission recommends no special dose
limits for women in general. Once pregnancy has been declared, the conceptus should be
protected by applying a supplementary equivalent dose limit, to the surface of the
woman’s abdomen, (lower trunk) of 2 mSv for the remainder period of pregnancy and
by limiting the intake of radio nuclides to about 1/20 (0.05) of ALI.
APPRENTICES AND STUDENTS
No occupational exposure is permitted below the age of 18 years. The
use of radiation by student below age of 18 should be discouraged. For students between
16 and 18years of age, the recommended limits for effective dose are 5 mSv, equivalent
dose to lens 50 mSv and to the skin or the extremities, 150 mSv. These doses are about
30% of the dose limits for occupational exposures for adults.
8 . 48
Table 6.1: Dose limits
Application
Annual Dose Limit
Occupational
Effective dose
Public
20 mSv per year,
1 mSv per year
averaged over a defined period
or
of 5 years, with no more than
1mSv/year
50 mSv in any single year @
averaged over
5 years
Annual equivalent Dose
Individual Organs
Eye Lens
150 mSv
15 mSv
Skin
500 mSv*
50 mSv
Hands and feet
500 mSv**
-
Equivalent dose
Pregnant Women
2 mSv for the surface of the abdomen
and 0.05ALI for intake of radio nuclides
after declaration of pregnancy up to the
termination of pregnancy
@
The limit prescribed by Atomic Energy Regulatory Board is 30mSv in a year
* Averaged over areas of no more than any 1 cm2 regardless of the area exposed. The
nominal depth is 7.0 mg cm-2.
** Averaged over areas of the skin not exceeding about 100 cm2.
DOSE LIMITS TO MEMBERS OF PUBLIC
Members of the public include children who might be subjected to an increased
risk and who might be exposed during the whole of their lifetime. In addition, they do not
make their own choice to be exposed and may receive no direct benefit from the
exposure. For planning purposes, it is considered appropriate to set limits for the
members of the public lower than those are for occupational workers. Hence, the limit for
public exposure is an effective dose of 1 mSv in a year. However, in special
circumstances, a higher value of effective dose could be allowed in a single year,
provided the average over 5 years does not exceed 1 mSv per year. For preventing
deterministic effects in lens of the eye and skin, annual dose limits of 15 mSv and 50
mSv respectively have been recommended (Table 6.1).
8 . 49
PERSPECTIVES ON OCCUPATIONAL EXPOSURE LEVELS
How well has the radiation protection system recommended by the ICRP has
achieved its twin objectives namely, eliminating the deterministic effects and reducing
the probability of stochastic effects to an acceptable level?
Table 6.2 shows for some important deterministic effects, acute threshold doses
(single exposure) and the annual dose rates, if received in highly fractionated or
protracted exposures for many years
The cumulative effect of which would be in excess of a threshold beyond which
clinical symptoms would be observable. These levels can be compared with the presently
recommended annual dose limit for whole body as well as specific organs given in the
last two columns. It can be seen that the dose limits are below the ones required for
elicitation of clinical symptoms of non-stochastic injury either by a single exposure or by
exposures year after year.
Further, the ALARA principle has assured that the exposures are generally kept
below the dose limits. As per the personnel monitoring data available with BARC, the
average annual exposures to all categories of radiation workers in India during the year
1998 is 1.49 mSv. This is about 1/30 of the maximum annual dose limit. As discussed
earlier, the risk of fatality from radiation induced cancer is 10-2 per Sv. On the basis, an
average annual exposure of 1.49 mSv would correspond to approximately 0.6 excess
fatality per year amongst the 40,000 radiation workers. This number is quite small
compared with, and indistinguishable from the total number of annual fatalities which is
nearly 800 (taking national annual fatality rate of 2%). Hence, it is evident that the
average exposure of a hypothetical member adds little to the hazards of every day living.
Radiation has always been a part of our daily life. We are constantly exposed to
radiation from both natural and artificial sources. Natural exposures come from the sun
and the stars (cosmic radiation), the earth’s minerals (uranium, radium, and rubidium)
and even from within our own body (40K,
14
C). As a result, the food we eat, the air we
breathe the water we drink and houses we live in, all contain traces of radioactivity.
Every inhabitant of the planet receives an average dose of 2.4 mSv per year (Table 6.3).
8 . 50
Table 6.2: Threshold doses for some deterministic effects of radiation for single (acute)
and protracted (chronic) exposures in comparison with annual dose-equivalent limits.
Tissue & Effects
Total Equivalent
Dose received in
a single brief
exposure (Sv)
Threshold Annual Dose
rate if received yearly in
highly fractionated or
protracted exposures
for many years (Sv/Y)
Temp. Sterility
0.15
0.4
Perm. Sterility
0.35
Annual dose limit for
Whole
Single
Body
organ
Expo.(Sv)
(Sv)
Testis :
Ovaries: Sterility
Lens
: Cataract
2.0
0.02
0.1
0.02
0.1
0.02
0.1
2.5 - 6.0
>0.2
5.0
>0.15
0.02
0.15
0.5
>0.4
0.02
0.17
1.5
>1
0.02
0.17
Bone marrow:
Depression of
Hematopoesis
Fatal aplasia
At any given place, however, the individual contribution varies depending upon
its geology, latitude, altitude, etc. Some of the coastal areas in Kerala have natural
background radiation levels as high as 10 mSv or more. The dose limit for
occupational workers is about 25 times higher than the normal natural
background radiation. But the average occupational dose is of the same order as the
natural background. Thus, it is to be emphasised that ionising radiation needs to be
handled with care rather than with fear and its risks should be kept in perspective with
other risks. The procedures available to control exposures to ionising radiation are
sufficient, if used properly, to ensure the radiation remains a minor component of the
spectrum of risks to which we are all exposed.
The most important source of artificial exposure (man made) is the use of
radiation in the medical field. More than 80% artificial radiation dose comes to people
from medical diagnostic procedures. The average background dose resulting from natural
radioactivity and man-made is estimated to be about 2-3 mSv. 87%of this is contributed
from natural radiation and remaining 13% from man-made sources. This is shown in
8 . 51
figure 6.1, a pie chart. The contribution from various components of the natural
radioactivity is shown in Table 3.
Table 6.3: Average World Wide Exposures to Natural Radiation Sources
Source of Exposure
Cosmic rays
Terrestrial Radiation
External
Outdoor
Indoor
Internal
Inhalation
Radon
Thoron
U,Th
Ingestion
K-40
U,Th etc.
Total
Annual Effective Dose
mSv
0.36
0.41
Annual Effective Dose
mSv
0.07
0.39
1.26
0.07
0.01
0.18
0.06
2.36
0.46
Fig 6.1: Percentage Contribution of Natural and Man Made Radiation to Background
--------------
8 . 52
7. RADIATION HAZARDS EVALUATION AND CONTROL
EVALUATION OF HAZARD DUE TO EXTERNAL RADIATION
External radiation (EH) hazards are caused when the radioactive source(s) or
waste lying near a work place or table (outside the human body). The chances of EH are
high from non-penetrating radiation such as X-rays and gamma rays on account of their
higher penetrating power. EH would less from beta and alpha radiation.
An estimate of EH due to non-penetrating radiation can be made by simple
calculations. It is particularly useful while planning radiological safety in the laboratory
deciding radiation shield and programming experiments with radiation sources. If one
knows the physical or chemical nature, strength of the source and the energy of the
radiation emitted, the safe radiation levels at working locations can be estimated. This
will further enable to project the approximate doses that might be received by a person in
general or typical operation(s).
Radiation level at any particular distance from a gamma source of given strength
can be calculated using specific gamma ray constant (k-factor) of that gamma ray
emitting source.
k-factor = 0.123 A.E mGy/hour (Specific Gamma Ray Constant)
This k-factor is constant for that gamma radiation emitting radionuclide and it is
the Exposure rate in mGy/h at ‘1’ metre from a point source of activity ‘A’ GBq
emitting gamma radiation of energy ‘E’ MeV. If the activity A is expressed in old unit of
activity i.e. in ‘Ci’ the exposure rate would be = 0.52 A.E R/h
SPECIFIC GAMMA RAY CONSTANT
(or exposure rate constant or k-factor)
Gamma emitting isotopes are characterized by a term known as specific
gamma ray constant (k-factor), which is defined as the exposure rate in
Roentgen per hour at 1 cm from the source of activity 1 mCi. These are shown
in Table 7.3. In SI units they are expressed in terms of mGy per hour at 1cm from 1MBq
of activity. Therefore, by knowing the strength of the source and its specific gamma ray
8 . 53
constant, exposure levels at any distance can be calculated. Conversely the strength of
the source can be estimated by measuring the exposure rate at a given distance.
CONTROL OF EXTERNAL HAZARD
EH can easily be controlled by adopting the three fundamental methods namely
(1) Distance (2) Time and (3) Shielding, depending on the situation. Some times only
one of the three methods would be adequate to follow to control the EH. But in certain
situation one may have to adopt all the 3 methods referred above.
Distance: Radiation exposure varies as Inverse Square of the distance from the source.
The variation of exposure rate with distance is given by the simple relationship.
I 1 D22

I 2 D12
where,
I1 = exposure rate at distance D1;
and,
I2 = exposure rate at distance D2
By knowing the exposure rate at a given distance D1, the exposure rate at any
other distance D2 can easily be calculated using equation referred above.
‘Larger the distance between radioactive source and radiation worker; lesser be
the radiation exposure rate’. It is, therefore, advisable to maintain optimum distance with
disturbing the main work with radiation source. To understand the application of inverse
square law following examples are discussed.
Example 1 : What would be the radiation level at 10 cm distance from a 10 MBq of
57
Co? Specific gamma ray constant for 57Co = 0.2 mGy/h (k-factor) at 1 cm distance.
Strength of the source (A) = 10 MBq
Exposure rate at 1 cm from to MBq source (I1) = 0.2 x 10 = 2mGy/h
Exposure rate at l0cm from the source = I2 = ?
I 1 D22

I 2 D12
Therefore


I 2  I1 D12  D22
Here I1 = 2.0 mGy/h, D1 = 1 cm, D2 = 10 cm
Therefore,
I 2  2 11  10 10 = 0.02 mGy/h
8 . 54
Example 2 : What is the distance required to reduce the radiation level from a 20 MBq
60
Co source to 0.025 mGy/h?
Exposure rate constant for 60Co = 3.1 mGy/h/MBq at 1cm
Strength of the source = 20 MBq
Exposure rate at 1 cm from the source = I1 = 20 x 3.1 = 62 mGy/h
Required exposure rate = I2 = 0.025 mGy/h
I 1 D22

I 2 D12
Here, I1 = 62 mGy/h, I2 = 0.025 mGy/h, D1 = 1 cm, D2 = ?
Substituting these values in the equation, we get,
D2 = 49.8 cm
Example 3 : While handling 25 mCi 65Zn source with a 15 cm tongs, within how much
time the operator will receive the weekly permissible equivalent dose. (Given k for 65Zn
= 2.7 R/h/mCi at 1 cm Assume 1 R = 1 rad).
Exposure level at 15 cm from 25 mCi Zn-65 source = 2.7  25  15 15
= 0.3 R/h
= 300 mR/h
Weekly permissible exposure
= 40 mrad
Allowable time
= 40  300 h
= 8 min
Time: Radiation dose received during work is directly proportional to the time
spent near the source. The dose received by a person remaining near a source for 10
minutes would be only half of that received by him in 20 minutes. Therefore, all the
radiation work should be planned in advance and should be executed in optimum
minimum time so that the exposure received could be kept as low as reasonably
achievable (Chapter 6). In order to minimize the time of operation with the actual
radiation source, it is advisable to perform trial operations with a dummy source. Source
should be taken out from the storage only when required or to be used.
Example 4 : In example- 3, if the user takes 4 min. to work with the source, the weekly
permissible exposure limit will be received in two such operations. Instead, if a 30 cm
8 . 55
tongs is used them 8 operation can be carried out without exceeding the weekly exposure
limit.
Radiation Shield: Where it is impracticable to adopt the method of distance and time
to ensure acceptable low radiation level at work place; method of radiation
shield is adopted. In this situation the radiation source(s) is/are shielded locally with an
adequate thickness of proper radiation shielding material. The type of radiation shield, the
material, and its thickness will depend upon the nature of radiation, strength of the
radiation source and the energy of radiation emitted. Alpha sources do not pose any
external radiation hazard because of their limited range in medium (even a few
centimeters of air can cut off most of energetic alpha particles emitted from radioisotope).
Therefore the alpha emitting sources do not need additional shield. But the beta emitting
radioisotopes need additional external shield to minimize the radiation level. These
radiation can travel far more distances in air than those of alpha radiation. In case of beta
particles, a fraction of energy, absorbed in the shielding material, will be converted to
bremsstrahlung (Chapter 2), which is similar to X-rays. The fraction of energy converted
in to bremsstrahlung depends on the atomic number (directly proportional to Z2) of the
shielding material and the energy of the beta particle (directly proportional to E). Hence,
to minimize the production of bremsstrahlung radiation from shielding material, it is
preferred to use a material of low atomic number such as perspex, aluminum, etc. A
perspex shield of 1 cm thickness would absorb most of the beta rays emitted by any
radioisotope. However, when a beta emitting radioisotope of very high activity (say in
curie) is to be shielded from radiation, a secondary shield of (Pb) lead is required to
absorb the small amount of bremsstrahlung produced in primary perspex shield.
X-rays and gamma radiation are penetrating radiation and can travel far off
distances in air, reducing intensity by inversely square principle. The radiation exposure
due to gamma radiation is brought down using high Z (atomic number) material as shield
such as Pb. Those materials are used as shielding for gamma where the process of
photoelectric effect is highest. The gamma radiation interact with shielding material by a)
Photoelectric, b) Compton and c) Pair production processes (Chapter 2). For medium
energetic gamma radiation, the photo electric effect in high in lead (Pb). Material of
8 . 56
higher Z such as tungsten, depleted uranium will be better shield than Pb for higher
energetic gamma radiation. Due to economic reasons concrete, brick, steel (low z
materials) are also used as structural shielding material to control radiation from X-rays
and gamma rays. The reduction in radiation level (Intensity or flux) of X and gamma rays
using shielding material is expressed by the following exponential equation:
I = Io e- .x
Where Io = original intensity of the beam,
x = thickness of the shield in cm
I = transmitted intensity through a shield (an absorber) of thickness ‘x’ cm
 = Linear attenuation coefficient of the shielding material (absorber)
 represents the fraction of the radiation intensity attenuated by unit thickness of the
material. It is expressed in units of cm-1.
Linear attenuation coefficient  will depend upon the energy of the radiation and
atomic number of the medium. Its value increases with increase in atomic number and
with decrease in energy of X and gamma rays. The exponential attenuation of radiation in
shielding material is shown schematically in figure 7.1. It can be represented either on a
linear or a semi-logarithmic graph as shown in figures 7.1 2a and 2b. Semi-logarithmic
plots are always straight line for an exponential relation. The slope of the graph gives the
attenuation coefficient (). These are similar to the exponential decay curves for
radioactivity. Practical application of  in above equation for calculation of intensity
evaluation, two terms introduced here a. Half Value Thickness (HVT) and Tenth
Value Thickness (TVT).
Half Value Thickness (HVT) : The half value thickness is that thickness of the
shielding material which will reduce the incident intensity of radiation beam to half. The
reduction factor offered by one HVT of the shielding material is 2 and by 2
HVT is 2 x 2 or 22. The reduction factor offered by ‘n’ HVT of shielding
material is 2n.
Tenth Value Thickness (TVT): Another specific thickness which is convenient to use
8 . 57
in shielding problems is the Tenth Value Thickness (TVT). TVT is that thickness which
will reduce the incident radiation intensity to one tenth. The reduction factor offered
by one TVT is 10, by two TVT is 10 x 10 or 102, by three TVT is 103 and so
on. Similarly the reduction factor offered by ‘n’ TVT is 10n.
For a given material both HVT and TVT depend upon the energy of the incident
radiation. The HVT and TVT values can be directly estimated from the attenuation
graphs figures 7.1 2a and 2b. The HVT, and TVT values for gamma rays of different
energy originated from different radio nuclides are given in table 7.3.
Fig. 7.1: Exposure attenuation. T½ = 2mm. (a) Linear Plot, (b) Semi logarithmic plot
(i.e. the vertical scale is logarithmic and the horizontal scale linear)
MONITORING OF EXTERNAL RADIATION
The adequacy of radiation safety provided by the above control measures will
have to be assessed by actual monitoring of radiation levels. Radiation monitoring
constitutes both area monitoring and personnel monitoring.
AREA MONITORING
The assessment of radiation levels at different locations in the vicinity of radiation
sources is generally known as area monitoring. Normally, area monitoring systems
should be able to monitor radiation levels in the range of 0.2 mR/h to 5 R/h (1.76 Gy/h
to 44 mGy/h) full scale and also have audio indication. A low range in the instrument is
8 . 58
useful in assessing radiation levels at occupied areas and general radiation survey around
installations. The most commonly used radiation monitoring instruments (e.g.
MINIRAD, MR4500) have miniature GM counters, making them useful over wide range
of exposure rates. MINIRAD monitor can measure up to 5 Rlh and MR4500 can
measure up to 50 Rlh making it useful in radiation emergencies. Both these instruments
can be used for area monitoring around source containers. Other monitoring instruments
in use are :
Beta-Gamma Exposure-rate Meter (SM-140:)It is an ionization chamber type survey
meter marketed by M/s. Electronics Corpn. of India Ltd. (ECIL), Hyderabad. It has
ionization chamber of size 400 cm3 and can measure X and gamma radiation exposurerate from 5 mR/h to 5 Rlh (corresponding to air kerma values from 0.044 mGy/h to 44
mGy/h) in three ranges. The chamber is provided with a window and by opening the
window beta radiations can also be monitored. This instrument is useful for general
purpose monitoring and for checking the radiation levels around radiation source
housings.
Radiation survey meter (MR-121): It is also marketed by M/s. ECIL, Hyderabad. It is a
GM type survey meter. It has a long glass walled GM counter and can measure X and
gamma exposure rates from 0.2 - 20 mR/h (corresponding to 0.88 μGy/h to 0.175 mGy/h)
in three ranges. It can also respond to high energy beta radiations. It is very useful for low
level area monitoring. This instrument however has a drawback that it does not respond
at high radiation levels.
Table 7.1 lists some of the radiation measuring instruments, their range, and
application. Two or more of these instruments are to be normally maintained so that in
case of failure of one, the other may serve as backup. As most of these instruments are
expensive and are likely to be used only occasionally, preventive maintenance and
periodic calibration checks are mandatory. In the case of ion-chamber based instruments,
preventive maintenance consists mainly of keeping the sensitive input circuitry free from
moisture by proper desiccation, removal of batteries when the instrument is not in use,
etc. For GM counter based instruments only removal of battery when not used for a long
period is required.
8 . 59
Table 7.1 : Radiation Monitors Commonly used in Radioisotope Laboratories
Name of the
instrument
MINIRAD
MR-4500
β-γ Exposure
Rate Meter
(SM-140)
RADMON
Type: RM 701
Radiation Survey
Meter
Gun Monitor
(Exposure Rate
Meter) GM-125
Radiation Survey
Meter PRM01215
GM Survey Meter
(MR-121)
DIGICON
Contamination
Monitor CM 710
Detector
Range
Use
Supplier*
GM counter
(4 ranges)
0-5 R/h
Pulsecho
system
- do Ionization
Chamber
0-50 R/h
0-5 R
(3 ranges)
Routine area
monitoring, Leakage
radiation level
around large
sources.
- do - do -
GM counter
0-2- R/h
(5 ranges)
- do -
- do -
0-5 R/h
- do -
Ionization
chamber
0-5R/h
- do -
Nucleonix
Systems Pvt.
Ltd.
PLA Electro
App. Pvt. Ltd.
- do -
NaI
scintillation
detector
0-1 mR/h
GM counter
0-20
mR/h
0-1000
cps
Background radiation
measurement: low
level radiation area
monitoring
- do -
ECIL
Beta/gamma
contamination
monitoring
Nucleonix
systems Pvt.
Ltd.
GM counter
ECIL
- do -
- do -
The simplest method of checking the performance of the instrument is to measure
the exposure rate, after it has been calibrated by the manufacturer, at a specific distance
from a source of known output and record the same for future reference (Refer Appendix
4). Performance checks can then be made at any time by comparing the recorded readings
with check readings made at the same distance from the reference source. While
comparing the reading with reference reading it is necessary to apply decay correction of
physical decay of radioactivity of reference source. If the check reading differs
considerably from the reference value, the instrument should be sent to manufacturer for
servicing and recalibration. In addition the operational and handling instructions should
be scrupulously observed to ensure prolonged and trouble-free performance of the
8 . 60
instrument.
INTERNAL HAZARD EVALUATION & CONTROL
Radioisotopes in unsealed form are handled in medical, industrial and research
institutions. When we talk about unseal sources, it means the sources are either in
gaseous, liquid, powder or paste form. The hazard due to handling of such sources can
result in both internal and external hazards. The severity of the hazard will depend
upon the quantity of radioisotope handled, the type of radiation emitted and its biological
behavior (metabolism) if enters the body system (leads to internal contamination). Unlike
external hazard, where the radiation will cease when the source is removed or the person
moves away from the source, the internal contamination will give continuous exposure to
body or organ where it gets deposited. The activity of internally deposited radioisotope
reduces by two means a) physical decay, and b) excretion through all excretion routes
such as urine, stool, sweat, sputum etc. The two means are known as the physical and
biological decay (or physical and biological half life) of the radioisotope.
Radioisotopes may gain entry into the body through any one of the following
pathways: a) Inhalation - by breathing contaminated air, b) Ingestion - eating drinking
food through contaminated hands or taking in contaminated food or water or transferring
radioactivity to the mouth through contaminated hands, c) Absorption of contamination
through intact skin or through wounds.
Control of Internal Hazard
Control of internal hazard is based on prevention of internal contamination
through anyone of the pathways (inhalation, ingestion or absorption) by blocking the
portals adopting several techniques such as containment of source, control of
environment, proper management of radioactive waste, use of protective
devices and good housekeeping,
1. Containment of Source: The radioactive source can be contained by using a tray
covered with polyethylene and absorbent sheets or paper or working in a Fume Hood
(FH) or Glove Box (GB). Where the radioactivity is being handled in the simplest
chemical or physical form in quantity kBq to few MBq, not exceeding 1 ALI . Fume
Hood - If there is a possibility of escape of activity in to the laboratory atmosphere (as
8 . 61
in the case of gas, vapor or aerosol and amount of radioactivity handled is between 1-10
times of ALI), it is necessary to use a FH to contain the activity. The purpose of FH is to
dilute the escaped activity with air and sweep out to release in to the outer atmosphere.
To facilitate this enough air is allowed to flow through the hood. The face velocity is kept
high to prevent the contaminated air escape from the face of the hood. The air flow in FH
is kept under negative pressure (i.e. the room air should enter to the FH and sucked out
through a pump and duct system. The exhaust fan is located at the end of the exhaust duct
line and all the duct work is also kept under negative pressure. Discharge duct of FH
should be independent and should not be coupled to other ventilation ducts. If the
radioactivity is an aerosol or gas, the escaping activity is filtered through appropriate gas
filters before releasing in to the atmosphere. Glove Box - If the nature of an operation
is dusty and complicated and the quantity is several times of ALI, the operations are done
in GB. The main function of GB is that it isolates the contaminant from the environment
by containing it to a enclosed volume. A negative pressure of at least 13 mm (0.5 inches)
water inside the glove box assures no leakage of air into the box.
2. Control of Environment: Environmental control includes proper design of
laboratory, evolving handling procedures for radioactive waste, isotope storage,
ventilation and direction of air flow, good housekeeping, decontamination of working
surfaces, floors and walls, regular monitoring of work surfaces and persons. The
following safety procedures are followed to handle unseal radioactive materials:
1. Work with unseal radionuclide should, as far as possible, be restricted to a
minimum number of rooms. This enables confinement and containment of
radioactivity.
2. Room surfaces should be smooth, easily cleanable without and sharp corners,
crevices, and angles.
3. Room ventilation should ensure that the air flow is from low to high activity
areas.
4. In some establishments, work with unsealed radioactive materials has to be
carries out in areas where non-radioactive work is also undertaken. In such
situations, the work with radioactive materials should be confined to designated
locations within the work area.
8 . 62
5. Protective devices such as hand gloves, lab coats and shoe covers (if necessary),
should be provided to the workers working in the lab. The main function of
protective devices is to intercept radioactivity that would otherwise contaminate
the worker's personal clothing and skin. Respiratory protection is required in
labs where there is likelihood of air contamination levels exceeding the
prescribed limits and is recommended only in special cases.
6. Barriers and change rooms should be maintained at locations (point of entry)
near radioactive area with facilities appropriate to each situation.
7. Sufficient space should be available since over crowding of working area
inevitably increases the potential for accidents.
8. No upholstered furniture should be allowed in the laboratory, metal furniture is
desirable as it is fire resistant and easy to decontaminate.
9. Operations should be planned to limit the spread or dispersal of radioactive
material. In general wet operations should be preferred to dry ones and transfer
of material should be minimised.
10. Equipment, glassware, tools, etc. should be identified for use in radioactive
area. All such items should be chosen with a view for easy decontamination or
for disposal. Special care should be taken to avoid contamination of costly
equipment for economic reasons.
11. Manipulations should be carried out over drip trays to minimise the effects of
breakage or spillage. Work surfaces should preferably be covered with an
impervious covering, which should in turn be covered with absorbent material
to contain spillage, such covers should be changed as frequently as needed to
prevent any accumulation of contamination.
12. Handling tools and equipment used should be placed in non-porous trays on
absorbent disposable paper which should be changed frequently. Pipettes,
stirring rods and similar equipment should never be placed directly on the work
surface.
13. After use, all vessels and tools should be set aside for special attention when
cleaning.
14. Mouth pipetting should not be done in laboratories handling unsealed
8 . 63
radioactive materials. Pro-pipettes may be used for dispensing radioactive
liquid. Smoking, eating, drinking and use of cosmetics should be strictly banned
inside the laboratory.
15. Procedures should be evolved for collection, segregation, storage, and disposal
of radioactive waste.
16. No person should leave the radioactive area without checking contamination on
hands, feet, personal clothing, and shoes. Suitable monitors (hand feet monitor)
may be made available at exit locations. Reference levels of contamination and
action to be taken in case the contamination found is more than the reference
level, should be displayed in the laboratory.
3. Contamination: In a laboratory, where unsealed isotopes are handled,
contamination of work surfaces or personnel may occur either from normal operations or
as a result of breakdown of protective measures. Contamination can be classified into two
categories transferable (loose) and non transferable (fixed). Fixed contamination
is the source of external hazard and is that which cannot be removed even after
repeated smearing. Loose or transferable contamination is that which can be removed by
easy smearing of surface and is the source of internal hazard. There are derived
permissible limits of contamination at specific surfaces for specific type of radiation.
These are listed in table 7.2. For surface contamination monitoring two methods are
usually followed. They are direct and indirect methods.
Direct method: In this method of monitoring, the counting instrument is directly placed
over the area for measurement of contamination. The contaminated surfaces are scanned
using portable contamination monitor to locate the area of contamination. By this
method, total contamination is both fixed and loose contamination is measured. This
method is useful at such locations where sampling by smear test is difficult. The
monitoring results obtained in this method have large degree of reliability, but is
unreliable when monitoring is done in an area having large background e.g. near an
isotope generator.
Indirect method: The indirect method is more suitable in case of loose or transferable
contamination. In indirect method of monitoring, the smear samples are taken and
counted separately in low background set up. For collection of smear approximately 100
8 . 64
cm2 surface area is marked and dry or wet swabs with filter paper or cotton is taken. By
measuring the radioactivity transferred to it, the level of loose contamination can be
evaluated. The reliability of monitoring results, in this method, is poor because the
quantity of radioactivity transferred to swab will depend on pressure applied, area
swabbed, moisture content in the swab, technique used etc. To evaluate the level of
contamination one should know before hand, the coefficient of removal of activity from
the surface. This coefficient is usually defined as the ratio of the radioactivity on the swab
to the actual radioactivity present on the surface. The dry swabs are less efficient than the
wet; the wet techniques cause partial penetration into deeper layers of the contaminated
surface. Indirect contamination monitoring is more suitable to detect contamination
caused by low energy beta radiation such as 3H and
14
C. It is also suitable to detect
contamination near high radiation area (e.g. storage of large amount of activity). The
detection limit of surface contamination by this method is usually one order of magnitude
lower than that the direct method. In many situations, both the methods are
complimentary and are used to obtain a complete picture of surface contamination. For
detection of alpha contamination ZnS (Ag) scintillation probe is used.
4. Air Contamination Monitoring: The assessment of air contamination is done by
using a static air sampler and filter paper as medium. Filter paper with good collection
efficiency (99.9%) is preferred. Glass fiber filter papers have about 99.9% collection
efficiency for 1micron size particulate. The volume of air to be sampled and sampling
time will depend upon the contamination levels in the laboratory. For assessment of
131
I
contamination in air, charcoal impregnated filter paper or ground charcoal is used. The
filter paper is then counted for collected radioactivity with the help of a pre-calibrated
counting setup. For measurement of 3H contamination in air, a different technique known
as cold strip method is used. The limits for air contamination on are governed by Derived
Air Concentration (DAC).
5. Personnel Contamination Monitoring: In practice it is difficult to correlate
contamination in a working place with the exposure to working staff. Therefore, the air
and/or surface contamination monitoring results are inadequate to assess the radiation
hazards. The complete environmental monitoring programme includes personal
monitoring also. It involves the monitoring of external (skin contamination) and internal
8 . 65
contamination in radiation workers. External contamination monitoring is done with an
appropriate portable or table type contamination monitors of known counting efficiency
for emitted radiations. For internal contamination monitoring, special techniques are
used. Most commonly used techniques are whole-body counting (for radio nuclides
emitting high energy gamma radiation), bioassay (urine analysis for 3H contamination) or
sometimes organ counting (thyroid counting in case of
131
I or
125
I). Personnel/individual
monitoring is carried out on routine basis or whenever it is necessary.
6. Decontamination Procedures: Removal of contaminant is the decontamination.
The procedure of decontamination should be such that it does not spread the
contamination to other areas. All efforts should be to contain the contamination to the site
of the incident. Whenever there is a need to move items out of the restricted area, they
should always be wrapped, to prevent spread of contamination. These considerations
apply equally to personal contamination and to contamination of equipment and facilities.
It is essential that decontamination is undertaken by properly informed and trained staff
only, using procedures and facilities that restrict doses to ALARA.
Personnel decontamination: The objective of personnel decontamination is to minimise
the radiation dose to the body and the skin. Decontamination techniques should remove
or reduce the external or internal contamination. External decontamination: Prior to
attempts to decontaminate externally deposited radioactive materials, all contaminated
clothing should be removed. Any skin swabs, nasal blows or other biological samples
should be retained in case they are needed to provide information in support of an
assessment of the incident. Contaminated clothing and skin swabs often provide a means
for the most rapid identification of radionuclide(s) causing the contamination. For injured
persons, first aid or partial decontamination is advised; only if there is likely to be delay
in reaching the medical facility. If contamination is localised over a limited body area
(e.g. hands, feet, neck, etc.) these areas should be cleaned by gently washing with soap
solution and water, but avoiding hot water. For more generalized contamination, the
person can be subjected to shower bath followed by a complete radiological survey.
Special attention should be given to body folds, the hair. Finger nails, but medical advice
should be sought before other decontamination methods are attempted. Care should
always be taken to avoid spreading contamination over the skin surface. Decontamination
8 . 66
should start on the outer perimeter of a contaminated area and proceed towards the
central point. Special care must be exercised to avoid spreading contamination to the
eyes, nose, or lips. Clean towels must be provided following each washing. Soaps, mild
detergents and approved chelating agents should generally suffice for skin
decontamination and these materials should be freely available in the workplace. In all
cases, washing must be gentle so that mechanical and chemical irritation of the skin is
minimised. Items and materials that have been used for personnel decontamination
should subsequently be treated as radioactive waste and treated accordingly. Special
medical attention should be sought if washing appears to be ineffective and the skin
contamination is not reduced to acceptable levels after several mild washings. Internal
decontamination: In case of internal contamination, the extent and magnitude of
contamination is unknown. Thus, detailed history of the accident or incident is often
important and must be recorded. The initial measurements and any samples obtained can
provide important evidence. Contamination that is caused internally (i.e., by inhalation,
ingestion or translocation from a wound or from skin) requires medical attention. There
are conventional methods, to be used under medical supervision for removing or
accelerating elimination of internally deposited radioactive material. Decontamination
of Surfaces: If the contamination is caused by short lived radio nuclides (e.g. 99Tc) the
decontamination by physical decay of activity is preferred. After a period of 8 to 10 half
lives the surface can be cleaned with moist swab. Water is most commonly used
decontaminating agent. If this attempt is unsuccessful, detergents like soap solution,
teepol, EDTA, etc. can be tried to remove the contaminant. These decontaminating agents
are very suitable for removing loose contamination from the smooth, hard, and nonporous surfaces. It is equally effective for those radioactive compounds which are water
soluble. To remove contamination from glass wares, the soap solution or chromic acid is
attempted. For the metal surfaces, where removal of contamination using above
mentioned reagents is difficult, mild acids like 0.1 N HCI or HNO3 may be tried. In all
these procedures, it is necessary to give time to moist the surface before removing the
contaminant. Fixed contamination of surfaces can be removed by gentle abrasion. If the
fixed contaminant emits high energy gamma radiation and has long half-life, the
contaminated object may be treated as radioactive waste and be disposed off
8 . 67
permanently. In case of spillage of large liquid activity in the laboratory, the contaminant
can be contained with in the area by placing enough absorbents carefully over it.
Protective tools like hand gloves, working clothes, shoe covers should be used.
Decontamination should be continued till the level of contamination reduced to the
specified levels. The specified levels are given in table 7.2.
INGESTION
INHALATION
WOUNDS
EXHALATION
GASTROINTESTINAL TRACT
RESPIRATORY
TRACT
PULMONARY
LYMPH NODES
BILE
LIVER
KIDNEY
URINE
BLOOD
SKIN
OTHER ORGANS
SWEAT
FECES
Fig. 7.2 : Schematic Representation of Routes of Entry, Metabolic Pathways and Possible
Bioassay Samples for internally deposited radio nuclides.
Table 7.2 : Derived Working Limits (DWL) for Radioactive Contamination
Location
Beta Bq/cme
Alpha Bq/cm2
Skin
1.5
1.0
Hands (Total)
350 Bq
250 Bq
Clothes
1. Personal
2
0.5
2. Plant
6
2
Shoes
1. Personal
0.37
0.037
2. Plant
6
2
Floor
3.7
0.37
8 . 68
Table 7.3 : Exposure Rate Constant for Certain Gamma Emitters
Isotope
Half
Life
Gamma Ray Energy (MeV)
Na
Na
42
K
46
Sc
52
Mn
2.6 y
15.0 h
12.5 h
84.0 d
5.7 d
57
267 d
60
5.27 y
245 d
60 d
8.06 d
1.274 (100%)
2.57 (100%) 1.37 (100%)
1.53 (18%)
1.119 (10%), 0.887 (100%)
1.43 (100%, 1.44 (6%), 1.34 (6%),
0.935 (84%), 0.744 (82%), 0.51
(58%)
0.136 (9% + 1% IC), 0.122 (89% +
1% IC), 0.0144 (6% + 84% IC)
1.33 (100%), 1.17 (100%)
1.14 (49%), 0.51 (3.4%)
0.035 (7% + 93 % IC)
0.36 (80% + 1% IC), 0.28 (5%), 0.08
(2% + 4 % IC), 0.72 (3%), 0.54 (9%)
0.662 (85% + 10% IC)
0.145 (49%)
0.317 (72%), 0.468 (47%), 0.308
(28%), 0.296 (26%), 0.604 (9%),
0.612 (6%), 0.589 (5%)
22
24
Co
Co
Zn
125
I
131
I
65
137
Cs
Ce
192
Ir
141
30 y
32.5 d
74.0 d
Exposure rate Constant
R/h/mCi mGy/h/MBq
at 1 cm
at 1 cm
12.0
2.8
18.4
4.4
01.4
0.33
10.9
2.6
18.6
4.4
00.9
0.2
13.2
02.7
00.7
02.2
3.1
0.64
0.17
0.52
03.3
00.35
04.80
0.78
0.083
1.1
RADIATION EMERGENCY AND PREPAREDNESS
Radiation emergencies in radioisotope laboratories would normally involve only
spillage of radioactive liquids. The preparedness and procedures to meet such
emergencies are briefed below.
Emergency Preparedness:
1. Charts, which detail various steps to be taken by a radiation worker, in case of a
radiation emergency, should be conspicuously displayed in the laboratory.
2. All the radiation monitoring and measuring instruments should be checked
routinely and kept always in working condition.
3. The ventilation system of the radioisotope laboratory should be checked
periodically and maintained properly.
4. A kit comprising of accessories like tongs, forceps, waste receptacles etc., which
are required for the decontamination operation should be available readily to
8 . 69
handle an accidental spillage.
5. A proper inventory of radioisotopes received, used, and disposed should be
maintained.
Emergency Procedures: It would be difficult to stipulate hard and fast rules for meeting
wide variety of radiation accidents. However, considering spillage of radioactivity as the
most likely accident in a radioisotope laboratory, following steps are recommended in
dealing with such emergencies.
1. Confine the spill immediately by tissue paper or such absorbent materials. This is
to avoid further spreading of the spillage.
2. Evacuate the immediate surroundings so that the spread of contamination by
accidental walking over the spill by laboratory personnel could be avoided.
3. If the spilled material has splashed on to a person or his clothing, immediate steps
should be taken to remove the contaminated clothes and to leave them in the area
meant for the purpose. The contaminated areas on the body should be washed
thoroughly with soap and water. Care should be taken not to scrub heavily or
inflame skin surfaces for removing the contamination. If internal contamination
has taken place, immediate action should be taken to minimize the deposit of
radioactivity in the internal organs and to enhance the excretion of the ingested
radioactive materials, under expert medical supervision. In case facilities are
available, bioassay or whole-body counting should be carried out to confirm
internal contamination.
4. Contaminated area should be decontaminated by experienced persons wearing
protective clothing like surgical gloves, shoe covers and surgical face mask etc.
tongs or forceps should be used to remove the contamination, which is confined
by absorbent materials. The absorbent materials so collected should be kept in a
polythene bag to be treated as radioactive waste. After as much contamination as
possible have been removed in this way, the surface should be-mopped with damp
(not wet) cotton or tissue paper held by forceps, always working' towards the
centre of the contaminated area, rather than away from it.
5. A contamination monitor should be used to monitor the area as well as the
personnel during the procedure of decontamination. The contamination monitor
8 . 70
should be operated by a person other than the person who does the cleaning up, to
avoid the possible contamination of the instrument. The contaminated gloves,
shoes covers, etc., should be kept in a polythene bag for decay for ultimate
disposal as radioactive waste. The forceps/tongs should be kept separately
covered in polythene bag for decay of radioactivity present on it.
6. If the spilled radioactive material is of a very short half life, say few hours, the
closure of the laboratory (if possible) for a day or overnight is recommended.
Since the natural decay process reduces the activity, the method is preferred to the
immediate decontamination steps as mentioned above. In this case, cleaning up
operations should be carried out once the activity is decayed to the safe levels.
7. In the case of large release of radioactive powder or aerosol, the room must be
immediately isolated from the surrounding by shutting off mechanical ventilation
and by closing windows and doors. A room with heavy air contamination can be
decontaminated from within by drawing the air of the room through an
appropriate filter.
8. Complete records of the accident giving details of the radioisotope, activity
involved and follow up procedures, etc. should be maintained.
For effective dealing with any kind of radiation emergency, the Radiological
Safety Officer (RSO) of the institution should educate and familiarize all the radiation
workers with the steps to be taken to meet the emergency.
Use of radio nuclides in research applications may not normally be a source for
exposure of general public, provided certain basic precautions are taken. These
precautions mainly involve good security for the radio nuclides, good work practice, and
a well controlled programme for the disposal of radioactive waste.
-----------
8 . 71
8. PLANNING OF RADIOISOTOPE LABORATORIES
INTRODUCTION
Handling of open radioactive materials requires a special radioisotope laboratory
depending on the nature of work. The radioisotope laboratory is graded in different
categories on the basis of the activities handled.
CLASSIFICATION OF RADIOISOTOPE LABORATORIES
The type of laboratory needed and the facilities to be provided for a radioisotope
laboratory using on radioisotopes (Le. liquid, powder or paste) would depend on a variety
of factors like (i) the type and level of activity to be handled (ii) chemical and
physical form of radioisotope and (iii) the typical procedure involved. Radio
nuclides are classified into four groups according to their relative radio-toxicity. Table
8.1 gives the classification of radio nuclides, i.e. high toxicity (Group I), upper
medium toxicity (Group II), medium and lower medium toxicity (Group III)
and low toxicity (Group IV). Further, based on the facilities available, three types of
laboratories namely, Type-I (simple), Type-II (medium) and Type-III (stringent) are
common (refer table 8.2). The quantities of radio nuclides that can be handled with
reasonable safety in different types of laboratories will also depend on whether operations
are i) normal chemical, ii) complex wet, iii) simple dry or iv) dry and dusty.
Table 8.3 gives the activities of radio nuclides of different groups that can be handled in
these three types of laboratories together with modifying factors according to type of
operation.
In order to illustrate the use of Table 8.1, 8.2 and 8.3, let us take a typical case. A
person is interested in using 59Fe, which is a gamma emitter. Referring table 8.1, it can be
seen that radionuclide belongs to Group III (medium and lower medium toxicity). Table
8.3 gives the prescribed limits for handling
59
Fe, depending on the type of operation, in
different types of laboratories, e.g. normal chemical operations, with less than 18.5 MBq
will need the facilities of Type-I laboratory and for more than 18.5 MBq, the facilities
required will be that of Type-II laboratory.
Further, if two or more radio nuclides from the same groups or different groups
8 . 72
are required, the quantities that can be handled in the laboratory should be adjusted so
that  Al/A*  1, where Al is the quantity of radio nuclides of the same or different
groups and A* is the prescribed limits for the classified laboratory depending on the
operation.
DESIGN OF AREAS FOR RADIOISOTOPE LABORATORIES
Having decided the type of laboratory required, one is interested to know about
the layouts of these laboratories. In designing a radioisotope laboratory, all aspects like
radiological safety, economy and convenience should be looked into. The immediate as
well as the future programme as regards the variety of radioisotopes and their
approximate quantities that might be handled should be considered.
GENERAL FEATURES
Site
The site of the radioisotope laboratory should be so chosen that there is no
increased background radiation level either due to operation of X-rays, tele-therapy units,
storages of radioisotopes, or due to handling of radioisotopes in the vicinity. It will be
advantageous to have all the rooms of the radioisotope laboratory grouped together,
preferably at one end of the building so that entry of persons not connected with that
work may be effectively prevented. Separate areas and separate facilities should be
provided for high and low level activity work to reduce the possibility of cross
contamination.
TYPICAL FLOOR PLANS
Sufficient area/rooms are required for each type of work, such as for (i) receiving
and storage of radioisotopes, (ii) preparation of radionuc1ides for application, (iii) actual
application of radio nuclides, (iv) segregation of subjects containing radio nuclides after
application, (v) sample counting and other measurements, (vi) temporary storage of
radioactive waste, (vii) decontamination and (vii) ground for disposal of radioactive
waste.
Other facilities will be required for non-radioactive operation such as utility and
st9rage room, a dark room, animal or plant room, offices and toilets, etc. Some typical
plans of radioisotope laboratories are given in Figures 8.1, 8.2 and 8.3.
8 . 73
Table 8.1 : classification of Isotopes
Basis – Relative Radio-toxicity per Unit Activity
GROUP I
210
Pb
230
Th
239
Pu
244Cm
210
Po
231
Pa
240
Pu
245Cm
223
Ra
230
U
241
Pu
246CM
226
228
Ra
233
U
241
Am
250C-
Ac
234
U
243
Am
252Cf
46
54
56
GROUP II
22
Na
90
Sr
125
Sb
137
Cs
182
Ta
228
Ac
Cl
91
Y
127m
Te
140
Ba
192
Ir
230
Pa
GROUP III
7
Be
41
A
52
Mn
65
Ni
C
K
56
Mn
64
Cu
F
K
52
Fe
65
Zn
Na
Ca
55
Fe
69m
Zn
75
82
85m
93
97
93m
129
131m
77
As
Y
97m
Tc
109
Pd
90
122
Sb
134
I
143
Ce
153
Sm
166
Ho
187
W
194
Ir
197m
Hg
220
Rn
36
14
42
Se
Y
97
Tc
105
Ag
92
125m
Te
135
I
142
Pr
152m
Eu
169
Er
183
Re
191
Pt
203
Hg
222
Rm
45
Ca
95
Zr
129m
Te
144
Ce
204
Tl
234
Th
18
43
Br
Y
99
Tc
111
Ag
127
Sc
106
Ru
125
I
152
Eu
207
Bi
236
U
Mn
110m
Ag
126
I
154
Eu
210
Bi
249
Bk
24
38
Cl
Sc
59
Fe
72
Ga
47
47
87
Kr
Zr
97
Ru
109
Cd
Te
135
Xe
143
Pr
155
Eu
Te
131
Cs
147
Nd
153
Gd
171
171
Er
186
Re
193
Pt
200
Tl
231
Th
227
Ra
232
U
242
Pu
249C-
Kr
Nb
97
Ru
109
Cd
Te
136
Cs
149
Nd
159
Gd
175
Tm
188
Re
197
Pt
201
Tl
233
Pa
Yb
185
Os
196
Au
202
Tl
239
Np
Co
115m
Cd
131
I
260
Tb
211
At
31
227
Th
237
Np
242
Cm
60
Co
114m
In
133
I
170
Tm
212
Pb
32
228
Th
Pu
243
Cm
238
89
Sr
Sb
134
Cs
181
Hf
224Ra
124
35
Si
Sc
57
Co
73
As
P
Vr
58
Co
74
As
S
Cr
63
Ni
76
As
87
85
Sr
Mo
105
Rh
113
Sn
Sr
Tc
103
Pd
125
Sn
48
Kr
Nb
105
Ru
115m
In
95
132
Te
131
Ba
147
Pm
165
Dy
177
48
99
130
I
140
La
149
Pm
166
Dy
181
51
91
96
132
I
Ce
151
Sm
141
185
Lu
191
Os
198
Au
203
Pb
W
193
Os
199
Au
206
Bi
W
Ir
197
Hg
122
Bi
69
71
Ge
Tc
147
Sm
238
U
Kr
Rh
187
Re
U-Nat
190
GROUP IV
3
H
Sr
113m
In
191m
Os
85m
15
O
Rb
129
I
193m
Pt
87
37
58m
91m
93
A
Y
131m
Xe
197m
Pt
Co
Zr
133
Xe
232
Th
59
Ni
Nb
134m
Cs
Th-Nat
97
8 . 74
Zn
Tc
135
Cs
235
U
96m
99m
85
103
Table.8.2 : Criteria for Grading Laboratories using Unsealed Radioisotopes
GENERAL:
Shielding (Against Gamma Radiation) and Availability of Qualified and
Trained Manpower as Required shall be ensured.
Type-I (Simple)
o A Simple chemical laboratory with good ventilation.
o Two rooms, one for handling and one for counting.
o Contamination monitor
o Ordinary Storage (with security)
o Sink-ordinary
o Table surface to be covered with smooth non-absorbent material.
o Remote handling tongs
o Pro pipette/remote pipettes
o Foot operated dust bins.
Type-II (Medium)
o Three or more rooms for storage, preparation, handling.
o Special table, floor and wall surfaces.
o Proper ventilation.
o Storage safe of concrete, steel or lead.
o Stainless steel sink (elbow or foot operated tap).
o Fume hood with special exhaust system.
o Contamination monitor
o Radiation Survey meter.
o Personnel Monitoring Badges.
o Planned radioactive waste disposal methods.
o Face masks,
o Glove box.
o Surgical Gloves.
o Remote handling tongs.
o Pro pipettes/remote pipettes.
o Foot operated dust bin
8 . 75
Type-III (Stringent)
Large scale laboratory-multi room complex with clear segregation of area based
on use, scale and type of operation with the radioisotopes. A general list is given below.
Contaminatrion monitor
Radiatrion survey meter
Air/alarm monitor
Foot, hand and clothing monitor
Pocket dosimeter
Personnel monitoring badges
Whole body counter
Bio-assay
Dilution and distribution room
Decontamination room
Stainless steel sink with elbow or foot operated tap
Respirators
Shoe barrier
Fume hood with absolute filter incorporated near junction of fume hood and
ventilation duct
Table 8.3 : Classification of Tracer Laboratories using Unsealed Sources
Group of
Radionuclide*
I
II
III & IV
Prescribed Limits for Handling Radionuclides
Type I
Type II
Type III
< 185 kBq
185 kBq-185 MBq
> 185 MBq
< 1.85 MBq 1.850 MBq-1.85 GBq
1.850 GBq
< 18.5 MBq 18.5 MBq-18.5 GBq
> 18.5 GBq
* group classification according to radio-toxicity.
Modifying Factors According to Type of Operation
1. Storage in closed, vented containers
x 10.0
2. NORMAL CHEMICAL OPERATIONS (e.g. analysis; simple
chemical preparations)
x 1.0
3. COMPLEX WET OPERATIONS (with risk of spills)
x 0.1
4. SIMPLE DRY OPERATIONS (e.g. manipulation of powders
and volatile radioactive compounds)
x 0.1
5. DRY AND DUSTY OPERATIONS (e.g. GRINDING)
x 0.01
10 .1
Ventilation
The flow of ventilating air in the radioisotope laboratory should be from area
having no activity to low activity area then to medium activity area and from there to area
of high activity. This flow of air pattern will prevent the spread. of radioactivity from
high activity to low activity area e.g. into the counting rooms or in the adjoining rooms in
the event of accidental spill.
The air in the radioisotope laboratory should not be re-circulated.
Surfaces
All surfaces in any active area should be smooth, unbroken and be made from
materials which are chemically inert, non-absorbent and water repellent. Consideration
should always be given to possible decontamination problems, which might arise, and so
materials must be chosen which are either easily decontaminated or which can be
conveniently removed or replaced.
Specifications
Areas/rooms where radioisotope laboratory is being planned, special design
consideration should be given for the following:
Walls, floor and ceiling
The basic requirement is that the walls, floor and ceiling should have a good clean
finish, which is free from cracks. From the point of cleanliness and ease of cleaning, it is
desirable to have coverings at angles of walls, ceiling and walls, and floor and walls. The
flooring should not be of any porous material like wood or concrete, in which case it will
be impossible to decontaminate after a spill. Asphalt tiles, rubber tiles: vinyl tiles or
linoleum have been found suitable for use as flooring materials, with the advantage that
contaminated sections can be removed and readily replaced. Cracks between squares can
be satisfactorily filled by heavy waxing of the surfaces. Further, floor should have at least
15 cm skirting on the walls.
Walls and ceiling are less likely to become contaminated than floor. Nevertheless
the surfaces should be smooth, crack free and non-porous. The porous wall material
should be coated with a non-porous, washable paint and preferably with a fine layer of
strippable coating.
10 .2
Work Surfaces
All work surfaces should have hard non-porous finishes which have the necessary
heat and chemical resisting properties such as Sunmica or Formica. It should be bonded
to the backing material with resin glue to give necessary temperature resistance. Further,
to ensure that the surfaces are not spoiled, it is advisable to have it always covered with
polythene sheet and then with absorbent paper, so that if there is any liquid spill. it is
immediately absorbed by the paper, which can be easily disposed off.
Containment System
To prevent spread of contamination into working areas, all work with
radioisotopes is carried out in containment systems like fume hoods/fume cupboards,
glove boxes, etc. The general purpose and the required degree of containment for
different systems are given below.
Fume hood
A fume hood is a box with sliding transparent front panel opening where an
inward air flow of 30-40 linear metres per minute is maintained. Fume hood is generally
used when contamination and external radiation hazard are not too high and relatively
low levels of activities are handled or operations where gases or reactions at elevated
temperatures are involved. Figure 8.4 presents a schematic drawing of a radiochemical
fume hood. A fume hood should not be placed near door ways or windows or in the
vicinity of strong air currents, which may tend to draw fumes from the hood. All the fume
hoods in anyone room should be controlled QY the same switch. This will avoid the
chance that air flow from an operating hood will bring contamination into the room from
a non-operating hood. The services generally required in a fume hood are water, gas,
vacuum and electricity. The controls of these services should be situated outside the hood
to minimize the number of movements through the front opening. The exhaust air should
be discharged so as not to contaminate surrounding facilities; a point of discharge 120150 cm above the roof or 120-150 cm above the tallest surrounding building is usually
satisfactory. Blower should be located near the top of the exhaust duct so as to minimize
the escape of active material caused by positive pressure within the duct. Usually it will
not be necessary to filter the hood output. If required, filters should be installed near the
10 .3
start of the exhaust duct.
Glove Box
A glove box is generally installed in active laboratories to facilitate the handling
of hazardous materials. It consists of a leak-tight enclosure in which objects or materials
may be manipulated through gauntlet gloves attached to ports in the walls of the box (see
Fig. 8.5). Its aim is to provide containment for materials which are either radioactive or
chemically toxic or both. Usually it does not provide any shielding protection against
penetrating radiation and so it is used for alpha or beta emitters. When gamma-emitting
isotopes have to be handled, a wall of lead bricks is usually constructed between the
operator and the glove box.
Glove box is maintained at a pressure slightly below that of the outside laboratory.
This means that air will flow into the glove box, should a leak develop and this will
prevent the contamination escaping. Two filters are normally placed on the ventilation
system-one to remove dust from the air being drawn into -the glove box and the second to
remove radioisotope particles from the air being drawn out of the box.
RADIOISOTOPE
STORAGE
STORAGE AREA
D
D
LIQUID WASTE
STORAGE
D
CM
WORK TABLE
D
CHANGE
ROOM
WB
WB
D
Fig. 8.1 : Typical Layout of Radioisotope Laboratory
10 .4
SINK
FUME HOOD
D
SOLID WASTE
STORAGE
FUME HOOD
ISOTOPE
STORAGE
COUNTING ROOM
SOURCE SEORAGE
AND
HANDLING
D
D
(a)
(b)
D
D
BETA
COUNTING
D
GAMMA
COUNTING
SOURCE STORAGE
AND
HANDLING
FUME
HOOD
ISOTOPE
STORAGE
Fig 8.2: Typical One room Radioisotope Laboratory
10 .5
FUME
HOOD
HHOT
GLOVE BOX
HOT ROOM
GAMMA LAB
D
D
GAMMA COUNTING
HARD BETA LAB
D
D
D
BETA COUNTING
MASS SPECTROMETER
AUTO RADIOGRAPHY
D
D
OFFICE
D
D
D
SOFT BETA LAB
WASH
D
D
Fig 8.3: Typical eight rooms Radioisotope Laboratory
10 .6
8
9
6
1
2
3
7
4
5
90 cm
165 cm
180 cm
1.
3.
5.
7.
9.
Hood Interior
Service Outlets
Controls for Service Outlets
Cabinets
Movable Sash
2.
4.
6.
8.
Fig. 8.4 : Radiochemical Fume hood
10 .7
Strippable Paint
Cup Sink
Electrical Outlets
Filter & exhaust
5
4
6
60 cm
3
2
1
90 cm
45 cm
7
8
1. Glass Window
3. Fluorescent Lamp
5. Filter
7. Air Lock
2 Service Inlets
4. Electric Power Panel
6. Exhaust outlet
8. Long Rubber gloves
Fig. 8.5 : Glove Box
SUMMARY
1. The radio nuclides/radioisotopes are classified into four groups according to their
relative radio-toxicity.
2. There are four Groups viz. Group I, Group II, Group III and Group IV. These
groups are called High toxicity, Upper medium toxicity, Medium toxicity and
Lower medium and low toxicity respectively.
10 .8
3. Depending on the facilities available for handling radio nuclides and the
quantities of radio nuclides of each group handled there are three types of
radioisotope laboratories. These are Type I, Type II and Type III.
4. Radioisotope laboratory should not be located in an increased background
radiation area i.e. dose to installation like X-ray, tele-therapy or accelerator, etc.
5. All the rooms of a radioisotope laboratory should be grouped together.
6. The flow of ventilating air in a radioisotope laboratory should be from areas
having no activity to low activity area; then to medium and from there to area of
high activity.
7. The air in the radioisotope laboratory should not be re-circulated.
8. All the surfaces of the radioisotope laboratory either working or any other should
be smooth, unbroken, non porous and of material which can be easily
decontaminated.
----------
10 .9
9. REGULATORY ASPECTS OF RADIOLOGICAL SAFETY
INTRODUCTION
Radiation sources are widely used for research purposes. Work with radiation
sources may involve exposure of workers to radiation. It is known that exposure levels
significantly greater than the limits prescribed by the ICRP may entail certain health
hazards. It is, hence, necessary to ensure that any work involving radiation exposure is
carried out in a manner that results in minimum radiation exposure to workers and public.
In order to have an effective control on the use of radiation and to ensure radiological
safety of the user as well as the public, the -Government of India has promulgated the
Radiation Protection Rules, 1971 under the Atomic Energy Act, 1962. The Act empowers
the Government to make rules relating to radiological safety. The responsibility of
enforcing the rules is assigned to the Competent Authority specifically appointed by the
Central Government. The Chairman, Atomic Energy Regulatory Board (AERB), is the
Competent Authority in India. All radiation protection requirements in respect of
institutions outside the Department of Atomic Energy are being enforced by Head,
Radiological Safety Division (RSD) of AERB.
In India, only persons who are duly authorized by AERB are permitted to procure
and handle radiation sources. If the sources are to be procured locally or to be imported
an Authorization/No Objection Certificate (NOC), should be obtained by the prospective
user from AERB.
CURRENT PROCEDURE
Application for authorization to handle radiation sources should be made to Head,
RSD, AERB. The applicant should furnish in the prescribed form details regarding the
type of sources, its activity, proposed use, names of the users, their qualification and
experience in the handling of radiation source, etc. (See APPENDIX -1).
RSD would scrutinize the application and evaluate the radiological safety status of
the laboratory and also would advise the applicant regarding the radiation safety
requirements, specific to his needs.
The advice would relate to:
10 .10
- safe storage of the sources,
- facility for handling the sources,
- laboratory fittings.
- personnel monitoring and area monitoring, - trained staff, and
- disposal of radioactive waste
- emergency procedures
Upon fulfilling the requisite safety requirements, the applicant will be authorized
by RPAD to procure and handle radioisotopes. . The above authorisation is liable to be
revoked in case of non-compliance of the safety regulations by the institution.
RADIOLOGICAL SAFETY OFFICER
The institution should appoint, subject to the approval by the Competent
Authority, a Radiological Safety Officer (RSO) whose duties and functions shall be as
follows:
a) to take all necessary steps aimed at ensuring that the operational limits of
radiation exposure to personnel are not normally exceeded.
b) to educate the radiation workers under his charge on the hazards of radiation and
on suitable safety measures and work practices aimed at minimizing exposure to
radiation.
c) to regulate the safe movement of all radioactive materials (including wastes
containing radioactive materials) within the area under his charge.
d) to maintain a log book for inventories like the name and amount of radioisotope
received, date of receipt, the department in which it will be used, date of using,
name of the user, etc.
e) to carry out periodic monitoring of areas and contamination check of surfaces
(wherever applicable) arid maintain these records in the log book.
f) to investigate and initiate prompt and suitable remedial measures in respect of any
situation that could lead to radiation hazards.
g) to ensure that reports of all hazardous situations along with details of any
immediate remedial measure that may have been initiated, are made available to
his employer for onward transmission to the Head, RSD, AERB, Niyamak
Bhawan, Anushaktinagar, Mumbai 400 094.
10 .11
h) to ensure that the ultimate disposal of waste containing radioactive materials is
done in a manner approved by the Competent Authority.
No person below the age of 18 years should be appointed as radiation worker. All
radiation workers should undergo a routine medical examination prior to employment
and thereafter "as specified by the Competent Authority from time to time.
The institution should maintain the following records in respect of its radiation
workers.
- Occupational history,
- Medical history, and
- Cumulative dose records.
SUMMARY
1. Government of India has promulgated the Radiation Protection Rules, 1971 (RPR,
1971) under the Atomic Energy Act, 1962.
2. The competent authority in India is Chairman, Atomic Energy Regulatory Board.
3. On behalf of the Competent Authority, the Radiological Safety Division RSD) of
AERB tenders advice to the users on radiological safety for ensuring the
regulatory requirements:
4. R.S.O. stands for Radiological Safety Officer.
5. Authorization/NOC for obtaining radiation sources is issued by RSD.
-----------
10 .12
10. DISPOSAL OF RADIOACTIVE WASTE
While working with radiation sources, some amount of waste inevitably arises.
Radioactivity reduces according to natural decay rates. As it decays there should be no
significant exposure of persons to radiation. The Central Government under Atomic
Energy Act, 1962 (33 of 1962, subsection of clause 30), has promulgated rules called The Atomic Energy (Safe disposal of radioactive wastes) Rules, 1987- for the safe
disposal of radioactive wastes.
The basic objective of waste management is to ensure that radiation exposure to
man and his environment do not exceed the prescribed limits.
There are two principal ways of managing radioactive wastes: i) storing under
controlled conditions permanently or until it has decayed to permissible limits or ii)
disposing into surroundings in such a way that natural processes do not transfer back to
human environment in amounts or concentrations to cause exposures greater than the
estimated limits. Storage implies intention to retrieve. Safety during storage to some
extent depends on surveillance. However, disposal means no intention to retrieve and
relinquishment of all controls and safety is assured by methods other than surveillance. In
some cases a combination of both the methods are used so that part of the radioactivity is
concentrated and stored while the remainder is discharged into the environment after
processing. In selecting the most appropriate method of disposal, the likely doses both to
the workers dealing with waste and to the members of public should be estimated.
The waste may be in a large variety of forms depending on the particular use to
which the radionuclide is put. These can be solid, liquid, gaseous, combustible or non
combustible,: aqueous or non aqueous. In chemical operations the waste may be in the
form of solutions, precipitates, contaminated equipment. In medicine or biological
research work the waste may consist of excreta, tissue specimen, foliage, or animal
carcasses. Here we will discuss the various types of wastes generated in a radioisotope
research laboratory with respect to their origin, treatment, and disposal
.
TYPES AND ORIGIN OF WASTES
In case of research laboratories the radioactive wastes can arise at a number of
10 .13
stages. In source storage room the sources are kept in shielded containers where
incidence of contamination is not very likely. However, if the primary container drops
out of a shielded container, it may cause contamination of the room as well as other
equipment owing to possible splashing. In source preparation and handling room the
incidence of contamination due to spilling on table tops, floor, and contamination of
hands, etc. are very likely. Decontamination of affected areas, objects and persons will
also generate solid and liquid radioactive wastes. In the case of administration to animals
in the form of injection the syringe itself will constitute solid radioactive waste. Solid and
liquid radioactive waste may arise in counting room also. Contaminated plants (foliage),
animals excreta, tissue specimens and animals carcasses which may be discarded after the
experiment all constitute solid waste.
Table 10.1 : Categories of
Liquid Wastes
Category
1
2
3
4
5
Activity level
A (MBqM-3)
A  10-10
10-10 < A  10-7
10-7 < A  1
10-3 < A  1
1<A
Table 10.2 : Categories of
Gaseous Wastes
Category
1
2
3
10 .14
Activity level
A (MBqM-3)
A  10-6
10-6 < A  10-2
10-2 < A
Table 10.3 : Categories 01 Solid Wastes
Category
1.
2.
3.
4.
Radiation dose on the surface of
wastes
D (mGy/h)
D  2 x 10-3
2 x 10-3 < D  2 x 10-2
2 x 10-2  D
 activity expressed in MBq/m3
Remarks
- emitters significant
 emitters insignificant
-emitters dominant -
emitters insignificant
CLASSIFICATION OF WASTES AND METHODS OF DISPOSAL
Radioactive waste can be classified in number of ways like high activity, medium activity
and low activity waste. However, to be quantitative, the International Atomic Energy Agency
(IAEA) has classified the liquid and gaseous wastes into various categories according to the
activity level and solid wastes according to the radiation dose at the surface. These are given in
Tables 10.1, 10.2 and 10.3.
In research laboratories we generally do not come across gaseous wastes. Table 10.2 is
given for completeness sake and will not be discussed further. The management of solid and
liquid wastes can be greatly simplified by segregating them into classes of material in such a way
that the constituents of any batch can be dealt with, in the same way e.g. separating solid from
liquid, high activity waste from low activity waste, long half-life from the short half-life,
combustible from non-combustible, aqueous solutions from organic liquids, etc. The guiding
principles for disposal of radioactive wastes are 1) concentrate and contain (2)
delay and discharge and (3) dilute and discharge. In selecting a method for
waste disposal, each radioisotope should be separately evaluated for hazard and it should be
borne in mind that radioisotopes discharged from different places may result in exposure of the
same individual.
DISPOSAL OF SHORT-LIVED SOLID AND LIQUID RADIOACTIVE
WASTE
Short-lived solid and liquid radioactive wastes having half life less than a year should be
disposed off in conformity with the recommendation given below.
Solid Waste
For the disposal of solid radioactive waste, segregation should begin at the point
of generation itself. Two waste containers for each work table must be used one marked
"active" and another "inactive". Solid radioactive waste should be accumulated in the
laboratory in suitable receptacles and subsequently buried locally in pits. Site for such burial
should be selected taking into account the topographical, hydrological, and meteorological
characteristics of the environment in order to minimize the impact of the waste on the
environment. Factors to be taken into account include also the nature and location of other
facilities in the vicinity, usage of ground and surface waters in the surrounding areas and the
anticipated risk of accidental dispersal of waste to nearby vegetation and subsoil waters.
The size of the pit may be 120 cm x 120 cm. The depth of the burial pit should be so
chosen that. the radioactive wastes have a top layer of compact earth of minimum 120 cm
thickness when the pit head is finally closed. Additional pits maybe made in the burial site, in
case the quantity of radioactive waste exceeds the amount permissible in one pit.
Successive pits should be separated by a distance of at least 180 cm. Not more than 12
pits should be made in one year. A log book should be maintained for recording the location,
identity, and quantity of each isotope buried in the pits. The burial site should be fenced off and
provided with a gate which should be kept locked. Placards should be displayed prohibiting
unauthorized entry.
It should be ensured that the total activity of the wastes containing radioisotopes buried at
any time and at anyone location (or a pit) of the burial site does not exceed the limits specified in
Table 10.4. However, when using more than one radioisotope at a time. the total quantity of
radioisotopes that can be buried should be calculated as follows: Determine the ratio. Ai/Li,
where Ai is the activity to 'be disposed and Li is the limit specified in table 10.4. The above ratio
summed over all the radioisotopes to be buried should not exceed unity i.e. 1.
A mixture containing 185 kBq of
131
I, 3.7 MBq of 3H and 37 MBq of
32
P is to be
disposed. Find out whether the mixture can be buried in a single pit at a time. If not, how it
should be disposed off?
Total quantity of radioisotopes (more than one) that can be buried at a time can be found
out from the equation Ai/Li  1
A1 for 131I = 0.185 MBq; L1 for 131I = 37.0 MBq; A2 for 3H = 3.7 MBq;
L2 for 3H= 9250.0 MBq; A3 for
32
P = 37.0 MBq; L3 for 32P = 370.0 MBq
Condition to be satisfied for a single pit buried is
A1 A2 A3


1
L1 L2 L3
Substituting the values
0.185 3.7
37


37.0 9250 370
This will work out to be <1, therefore the mixture can be disposed at a time in a single
pit.
Normally after a period of seven to ten half-lives the radioactivity would decay
sufficiently so as to permit the contents of a pit to be removed and disposed off into the
municipal dump as normal waste. The aforesaid pit may then the reused for further burial of
radioactive waste. However, the waste should be released only after careful monitoring.
Liquid radioactive wastes
Short-lived liquid radioactive waste may be disposed off in the sanitary sewage system,
provided that the quantity is small. All radioactive wastes so released must be soluble
or dispersible in water. The quantity of the waste should not exceed the limits specified in
Table 10.5. It is assumed that the outflow of non-active effluents from other parts of the institution is such that it will dilute the radioactivity by a factor of about 100 before reaching the
main sewage. The total activity of all radioactive waste discharged into the sanitary sewage in
one year should not exceed 37 GBq (1 Curie). Disposal of waste containing 3H and
14
C
though long lived may also be done on similar lines subject to the limits given in table 10.5.
If more than one radioisotope is present in the waste and if the identity and activity of
each isotope is known, then the limiting value for disposal should be derived as discussed above.
If the identity of each radionuclide in the mixture is known but the activity of one or more
radio nuclides in the mixture is not known, the total activity limit for the mixture is the limit
specified in the table 10.5 for the radionuclide having the lowest limit. A log book should be
maintained for: recording the identity and quantity of each radioisotope and the time of its
disposal into the sanitary sewage system.
Used liquid scintillation cocktail should not be disposed off in
sanitary sewage system. It should be treated as a chemical waste.
Disposal of animal carcasses
For disposal of radioactive carcasses, wrap each animal in a separate polythene bag
ensuring that the claws are duly covered so that they do not puncture the bag. These bags may be
buried in pits as per the procedure for solid radioactive wastes.
.
Disposal of radioactive foliage
The size of radioactive foliage may be reduced if possible (by drying and mechanical
compression) any wrapped carefully in polythene bags. These bags should be buried in pits as
per the procedure for disposal of solid radioactive waste.
Table 10.4: Disposal Limits for Ground Burial
Radionuclide
3
H
14
C
24
Na
32
P
35
S
36
Cl
45
Ca
60
Co
85
Kr
59
Fe
89
Sr
90
Sr+ 90Y
94
Zr + 95Nb
Maximum activity
in a pit (MBq)s
9250
1850
370
370
1850
37
370
3700
3700
370
37
3.7
370
Radionuclide
99
Mo
Ru + 106Rh
124
Sb
125
I
131
I
131
Xe
137
Cs + 137m Ba
144
Ce + 144Pr
170
Tm
192
Ir
210
Po
106
Maximum activity
in a pit (MBq)
370
37
37
37
37
37
37
37
370
370
3.7
INCINERATION
Incineration of combustible waste (such as foliage, crops carcasses, etc.) releases part of
the radioactivity to the atmosphere and the remaining retained in ash. However, incineration
should be done under controlled conditions to ensure that gaseous radioactivity does not affect
the immediate environment and the ashes collected for separate disposal as solid waste. Since
these are difficult to achieve in practice, incineration is not generally recommended.
In case an incinerator facility is available, radioactive waste should be incinerated in a
separate batch and the ash collected by wet method separately for further pit-disposal.
DISPOSAL OF LONG LIVED AND IN-DISPERSIBLE RADIOACTIVE
WASTES
Solid and liquid radioactive wastes containing long lived radioisotopes (half-.life of the
order of years), (except 3H,
14
C) and in-dispersible wastes should not be locally disposed off.
Advice regarding the safe procedure and permission for packing, safe transport and disposal of
radioactive wastes should be obtained by the user from the Head, RSD, AERB.
Table 10.5 : Disposal Limits for Sanitary Sewage System
Radionuclide
3
H
C
24
Na
32
P
35
S
36
CL
45
Ca
60
Co
89
Sr
90
Sr + 90Y
94
Zr + 95Nb
99
Mo + 99m Tc
106
Ru + 106Rh
124
Sb
125
I
131
I
137
Cs + 137m Ba
140
Ba + 140La
144
Ce
170
Tm
192
Ir
210
Po
14
Maximum limit on total
discharge per day
(MBq)
92.5
18.5
3.7
3.7
18.5
0.37
3.7
0.37
0.37
0.037
3.7
3.7
0.37
0.37
3.7
3.7
3.7
0.37
0.37
3.7
3.7
0.037
Average monthly concentration of
radioactivity in the discharge
(MBqM-3)
3700
740
222
18.5
74
74
10.1
37.0
11.1
0.148
74
185
14.8
25.9
22.2
22.2
185
29.6
11.1
37.0
37.0
0.74
Summary
1. For the safe disposal of radioactive wastes, the Central Government, under Atomic Energy Act
1962 (33 of 1962, subsection of clause 30) has promulgated rules and these are called, The
Atomic Energy (safe disposal of radioactive wastes) Rules, 1987.
2. Solid and liquid wastes should be collected separately in suitable containers in the laboratory
itself for disposal.
3. Three guiding principles for the disposal of radioactive wastes are (i) concentrate and contain, (ii)
delay and discharge and (iii) dilute and discharge.
4. The basic objective of waste. management procedure is to ensure that it will not
result
in
exposure to man and his environment by more than the prescribed limits.
5. Short lived liquid radioactive waste, soluble or dispersible in water (including those containing
3
H and
14
C) should only be disposed off in sanitary sewage system. Organic waste from liquid
scintillation counting should-not be disposed off in sanitary sewage system. It should be treated
as chemical waste.
6. Short-lived solid radioactive wastes should be disposed off by burying in pits.
7. If separate incineration facility is used for the burning of combustible radioactive wastes, the ash
should be collected by wet method and disposed off as solid radioactive waste.
8. Any advice-regarding waste disposal should be obtained from the Head, RSD, AERB.
-----------
11. TRANSPORTATION OF RADIOISOTOPES
INTRODUCTION
On receiving a purchase order from an authorized user, radioisotopes are packed, marked,
labelled by the supplier and then sent to the consignee by rail, road, sea or air. During transport it
is necessary to ensure both the security of the consignment and further the exposure to the
transport operators and other members of the public should be kept to a minimum. For this
purpose transport of radioactive materials is permitted only in packages of prescribed design as
governed by national/international regulations.
DESCRIPTION OF PACKAGES
A package should have a containment system for the purpose of housing the radioactive
material during transport. It should be provided with a closure device (e.g. lid) in order to
prevent release of the source during transport. It should also incorporate adequate shielding. The
maximum permitted radiation level at the external surface of a package is 2 mSv/h (200 mrem/h)
and the corresponding limit at 1 m from the external surface of the package is 0.1 mSv/h (10
mrem/h). Accordingly they are classified in terms of Type of package (depending upon the
sturdiness of the packaging) and Category of package (depending upon the radiation level
outside the package).
TYPES OF PACKAGES
The various types of packages are a) Excepted packages, b) Low specific activity (LSA)
packages, c) Surface contaminated objects (SCO) packages, d) ‘Type A’ packages, e) ‘Type B’
packages. An excepted package is either an empty package having contained radioactive material
or one which contains very small quantities of radioactive material prescribed in relevant
regulations as accepted from the regulatory requirements. If the activity of the radioactive
material per unit mass or unit volume is low enough, or if the material to be transported is a
contaminated object such that the contamination per unit surface area is within a specified, limit,
such material is described as low specific activity (LSA) material or surface contaminated object
(SCO), as appropriate. These consignments pose limited radiological hazard as could be
imagined from the above description. ‘Type A’ and ‘Type B’ packages can contain higher
amounts of radioactive material. A ‘Type A’ package is one which is designed to withstand
normal conditions of transport. ‘Type B’ packages should be sturdier and have to be designed to
withstand severe accidents that may occur during transport. These include crash, fire, and
immersion in water. Each package has to be type tested before approval for transport. The tests
simulating normal and accident conditions of transport are specified in the regulations for safe
transport of radioactive material.
CATEGORIES OF PACKAGES
Packages are classified into three categories. Before going into categories of packages let
us first define an important quantity, viz., the transport index. Transport index is a number
expressing the maximum radiation level in mrem/h at 1m from the external surface of the
package.
Category I WHITE: If the radiation level at any location on the external surface of the
package does not exceed 0.5 mrem/h and if the transport index of the package is 0.0 then the
package belongs to Category I WHITE.
Category II YELLOW: If either of the limits for Category I WHITE is exceeded then the
package belongs to Category II YELLOW, provided, however, that the radiation level at any
location on the external surface of the package does not exceed 50 mrem/h and the transport
index of the package does not exceed 1.0.
Category III YELLOW: If either of the limits for Category II YELLOW is exceeded then the
package belongs to Category III YELLOW, provided, that the radiation level at any location on
the external surface of the package does not exceed 200 mrem/h and the transport index does not
exceed 10.0. However, if a package is transported under exclusive use conditions (i.e. chartered
vehicle) the radiation level at the external surface of the package shall not exceed 1000 mrem/h.
On Receipt of Packages
The responsibility for safety in the transport of radioactive material rests with the
consignor. However the procedure to be observed while receiving radioactive consignments is
briefly described below:
1. Immediately upon receipt of intimation from the carrier regarding the arrival of the
consignment, arrangements should be made to take delivery of the package.
2. Only persons who are familiar with radiological safety should be sent to collect the
package.
3. The person who collects the package should wear his personnel monitoring badge. He
should also carry with him a polythene bag in order to house the package in case it is
received in a damaged condition.
4. Before collecting the package it should be ensured that the package is indeed addressed to
the institution on whose behalf the package is being collected.
5. If the package is in a damaged condition the observed condition of the package should be
recorded in the documents of the carrier and the package should be taken delivery of,
carefully wrapped in the polythene bag brought for the purpose and taken to the
institution. The matter should be immediately reported to the supplier and also to Head,
RSD, AERB, Niyamak Bhawan, Anushaktinagar, Mumbai - 400 094 by telephone,
telegram or telex.
12. PRODUCTION OF RADIOISOTOPES AND LABELLED
COMPOUNDS
Radioisotopes are used today in almost every field of human endeavour. Its use in
industry, medicine, agriculture, and research represent some of the beneficial applications of
atomic energy. The application of radioisotope can be of simple irradiation of samples like in
industrial radiography, sterilization of medical products, irradiation of food items, radiotherapy
etc. or as tracers like in nuclear medicine, hydrology, mining etc. A number of radioisotopes
occur naturally and a large variety of them are produced artificially in nuclear reactors or by
particle accelerators. About 200 radioisotopes are currently available. In the production of
radioisotopes by artificial means, a suitable projectile is made to bombard a suitable target
element (or compound). The projectiles generally used are neutrons, deuterons, alpha particles
and some light nuclei. All the presently available radioisotopes in the country are produced by
neutron bombardment in nuclear reactor.
The most frequently occurring reactions during neutron bombardment of the target are of
the following types:
a. A (n, ) B reaction
59
Co + 1n
60
Co
The product nuclide of this type of reaction is an isotope of the target element.
b. A (n,) B
C reaction
130
Te + 1n
131
I + - + 
In this reaction the radioactive daughter element C is produced by the disintegration of
primary product B. This type of reaction is useful in the preparation of carrier-free materials
c. A (n,p) B reaction
35
Cl + 1n
35
S + 1P
32
32
S + 1n
P + 1P
The product nuclide differs chemically from the target element.
d. A (n,) B reaction
6
Li + 1n
3
H + 4He
In this type of reaction also, the product nuclide is, chemically different from the target
nuclide.
e. Fission Products
When Uranium-235 or Plutonium-239 absorbs a neutron, fission of the nucleus takes place
to a variety of fission product nuclides of which many are radioactive.
235
U (n,f) fission products like 137Cs, 99Mo, 147Pm, etc.
The amount of radioactive atoms (Activity) produced is dependent on the number of
atoms of target, neutron flux of the reactor, neutron cross section, time of irradiation and decay
constant of the product radionuclide.
SEPARATION OF ISOTOPES: CHEMICAL PROCESSING
When a radioisotope is to be simply used as a radiation source, the irradiated target
material itself is used directly or after proper sealing in a suitable container. Iridium-192, Cobalt60, Thulium-170 sources are some example of this type. However, if there is a need for an
isotope free from other radioactive impurities in a specified chemical form, it is necessary to
carry out chemical processing of the irradiated target material.
The most generally used separation methods for isotope processing are filtration, co
precipitation,
ion-exchange,
solvent
extraction,
distillation,
vacuum
sublimation
and
electrochemical technique.
The isotope thus produced can be a) Carrier free, i.e., it does not contain any other
isotopic impurity. e.g.,131I free from all other isotope of I, b) no carrier added, i.e., it contains
some of the isotopes of the intended element, produced during the production but no isotope is
added during separation or purification process externally, c) Carrier added, i.e., isotope is
added during the separation or purification process externally.
Production of labelled compounds
A labelled compound is defined as a molecule in which one of the atoms is replaced by
its isotope, which can either be a stable or a radioactive one. It is also known as a labelled
molecule, a tagged compound, a radiotracer compound or simply a radiochemical (generally the
radiochemical term is used for simple chemical forms like halides, oxides, phosphates etc.) A
radio labelled compound can either be, a) specifically labelled, b) uniformly labelled, c)
randomly labelled or d) non-isotopic labelled or foreign labelled. Some examples of such
labelled compounds are given below:
a) Specifically labelled
CH3CH(CH3)CH2CH(NH2)COOH
Leucine (1-14C)
here carbon at position 1 is 14C.
b. Uniformly labelled
CH3CH(CH3)CH2CT(NH2)COOH
Leucine (U-14C)
here all the carbon atoms are 14C
c. Randomly labelled
CH2TCT(CH3)CH2CT(NH2)COOH
Leucine T -(G)
here tritium label is randomly positioned
d. Non-isotopically or foreign labelled
Protein labelled with 125I; Nucleotide labelled with 35S.
(In these cases. the isotope used for labelling is not an isotope of any of the elements naturally
present in the molecule).
Carbon and hydrogen being present in almost every organic compound, Carbon-14 and
tritium labelled compounds are being used extensively. In recent years, phosphorus-32, sulfur-35
and
125
I labelled compounds are finding wide applications in the frontier areas of research in
genetic engineering. All these labelled compounds are synthesised either by chemical or
biochemical methods.
Specific Activity of Radio-labelled Compounds
The amount of radioactivity in a labelled compound is expressed in terms of specific
activity, which is the total activity divided by free quantity of the element, chemical or labelled
compound.
S = A/W , where A is activity and W, the weight in grams or moles.
This is an important parameter, which should essentially be known for any application,
which involves quantitative analysis. .
Specific activity is expressed in units of Ci/g, mCi/mmol, mci/ng or MBq/mmol.
Storage of Radio-labelled Compounds
Radio-labelled compounds unlike their non-radioactive counter parts undergo radiolytic
decomposition, the extent of which depends on the half-life of the isotope, the specific activity,
type of radiation and its energy and the nature of the compound. To minimise radiolytic
decomposition, the labelled compound must be stored in i) an inert medium, ii) as dilute
solutions or solids iii) in the presence of radical scavengers such as ethyl alcohol, iv) in the
absence of oxygen or v) at low temperatures ranging from 4 to 183°C depending upon the nature
of the compound
Quality and Purity of Radio-labelled Compounds
The labelled compounds have to be checked for their identity and chemical and
radiochemical purity. This is the most important aspect, where quality specifications are defined
depending on the nature of intended use. Every radio-labelled compound has to undergo
following tests.
a. Physical form [solid, liquid(solution)] or gas and colour
b. Radioactive content (the chemical form and activity)
c. Specific activity (activity per unit weight)
d. Radioactive concentration (activity/ml)
e. Radiochemical purity (the percentage of activity in the stated chemical form)
f. Radionuclide purity (the percentage of activity of the stated radionuclide)
Then depending on the intended end use, the tests for chemical purity, biological purity and
integrity of source/package etc., are carried out. For this purpose, analytical techniques like
chromatography (HPLC, Reverse Phase Chromatography etc.) spectroscopy like UV, IR, NMR,
Gamma spectroscopy etc., are employed.
Table 12.1 : Some important Radioisotopes
Isotope
Method of Production
Half life
Decay mode
14
C
3
H
137
Cs
14
N(n, p) 14C
6
Li(n, ) 3H
235
U (n, f)
5568y
12y
30.17y


, 
11
11
13
12
20.4m
9.96m
122s
13h
+
+
+
EC
Major
Application
Industry
Medicine
Radiation
Sources
Medicine
Daughter Te99m is used in
medicine
Radiation
Sources
Research
Research
Radiation
Sources
Medicine
Research
Medicine
Medicine
24
Na
51
Cr
6Co
23
Na (n, )24Na
50
Cr (n, ) 51Cr
59
Co(n, ) 60Co
15.06 h
27.8 d
5.2y
, 
ec, 
, 
131
I
99
Mo
130
Te(n, ) 131Te -131I
98
Mo(n, )99Mo
235
U (n, f)
8d
67h
, 
, 
192
191
74.37d
, 
9.74m
110m
EC
+
Medicine
Medicine
Ir
C
N
15
O
123
I
62
Cu
F
18
Ir(n, ) 192Ir
B(p, n) 11C
C(d, n) 13N
14
N(d, n) 15O
124
Te(p,2n) 123I
127
I(p,5n) 123Xe EC 123I
62
Ni(p,n) 62Cu
18
O(p,n) 18F
---------
APPENDIX -1
PROCEDURES FOR ESTABLISHING A RADIOISOTOPE LABORATORY AND FOR THE
PROCUREMENT OF RADIONUCLIDES
Radioisotopes in India can be procured and handled only by users who have either
worked under an authorised person or undergone training in safety and are duly authorised by
Radiological Safety Division (RSD), Atomic Energy Regulatory Board (AERB), Niyamak
Bhawan, Anushaktinagar, Mumbai - 400 094. This authorisation is based on the radiological
safety status of the laboratory, where radiation sources are handled. For this authorisation, it is
mandatory that an application is submitted to RSD, furnishing the details of the laboratory and
qualifications and experience of personnel, in the prescribed format available at web site of
AERB (www.aerb.gov.in) along with the plan of the radioisotope laboratory.
The plan of a radioisotope laboratory depends upon the type of radioactive material to be
handled, its physical form and activity and the technique of its application in the study. In
general the minimum requirement is of two rooms of suitable size (adjacent to each other) one
for storage, and handling of radioisotopes and the other for counting of radioactive samples. A
second counting room may be provided taking into account future expansion. This will ensure
separate rooms for ‘high activity’ and ‘low activity’ counting. If experiments with animals are to
be carried out, then a separate room for segregating ‘active’ animals is to be provided. This will
imply that the area will depend on the activities planned to be handled.
Two copies of the plan of the radioisotope laboratory, drawn to scale (1:50) should be
sent to RSD, indicating in it the room for storage and handling of radioisotopes, counting
radioactive samples and keeping animals etc. The dimensions of the rooms, the position of doors,
windows, exhausts, fume hoods, work benches and other fixtures etc. also should be indicated in
the plan. Preferably a site plan (to scale) of the building housing the radioisotope laboratory may
also be sent along with the plan of the laboratory indicating the nature of occupancies in the
immediate surrounding of the radioisotope laboratory including those above the ceiling and
below the floor, if any.
The suitability of the plan is generally decided on the basis of the above information and
if found necessary requisite modifications are suggested by RSD before approval. Otherwise one
of the plans of the radioisotope laboratory is returned duly stamped and approved.
Order for the requisite quantity of radionuclides is placed with Senior Manager,
Technical, Sales and Operation, Board of Radiation and Isotope Technology (BRIT), V.N. Purav
Marg, Deonar, Mumbai - 400 094 by the user. This will be referred to RSD by BRIT for
Authorisation/No Objection Certificate. Queries regarding approval and classification of
radioisotope laboratory, approval of qualified staff, nomination of RSO, Authorisation/No
Objection Certificate are to be referred to the Head, RSD, whereas queries regarding the supply
of radionuclides are referred to BRIT.
APPENDIX - 2
RADIATION PROTECTION SURVEY OF A RADIOISOTOPE LABORATORY
Contamination of work places is a potential hazard in the use of radionuclides in open
form in a radioisotope research laboratory. It is therefore mandatory that area monitoring and
contamination check is carried out twice a month and record is maintain by a trained person or a
Radiological Safety Officer (RSO). Any unusual occurrence should be reported to Head,
Radiological Safety Division (RSD), AERB, Niyamak Bhawan, Anushaktinagar, Mumbai - 400
094 immediately. The consolidated reports of all surveys should be sent to Head, RSD in the
month of March every year.
During the area survey, radiation levels must be determined at the source storage, waste
container, preparation room, counting room, work place, etc. Since radiation levels expected are
normally low, the monitoring instrument should be properly selected (e.g. GM survey meter).
During the survey, the instrument should be kept at the highest sensitivity range and held close to
the surface being monitored and moved across slowly to cover the required area. If the radiation
level exceeds the range, the meter should be switched to the next higher range. The radiation
levels at various places should be recorded on a layout plan of the room. It should be noted that a
background radiation level exists at all places and may range from 10-20 R/h. The radiation
level at any place occupied by radiation workers should not exceed 2.5 mR/h and one tenth of
this for places occupied by non-radiation workers. All attempts must be made to keep the levels
further down, as low as reasonably achievable. During the survey it should be ensured that all
areas such as storage facility, waste container, work place, etc. have labels indicating radioactive
area displayed at the appropriate place.
Contamination checks must be carried out with suitable instruments (e.g. , , γ
contamination monitor) at all susceptible places:, Some of these are work area. preparation room,
wash basin, door knobs, etc. In addition, hand and clothing should also be monitored. The
procedure would involve moving the detector (with window open), slowly, over the surface to be
monitored and recording of the readings. It should be noted that there will always be some
background level of counts in the range of a few counts per minute depending' upon the
equipment, place, etc. Activity above background level should be considered as contamination.
Contaminated materials should be kept aside either for decontamination (see Appendix 3) or as
radioactive waste for disposal.
Contamination by low energy beta emitters such as 3H, 14C, 35S cannot be detected by
GM counters. For this purpose a swab test must be carried out as described in Appendix 3.
Results of the radiation protection survey must be recorded in the attached proforma.
Proforma for Radiation Protection Survey of a Radioisotope Laboratory
1. Name of the institution
:
2. Name and designation of
head of the institution
:
3. Address of the institution
:
Telephone No.
Telex No.
:
:
4. Name of the department in which :
radioisotope laboratory is situated
5. Nature of work carried out
:
(example : normal chemical
operation, analysis, tracer studies,
simple wet operation, simple dry
operation, complex wet operation,
dry & dusty operation, etc.)
B. Staff
Name
1. Head of the department
:
2. Radiation Safety Officer
:
Approved by AERB
:
3. Details of other radiation workers
Sr.
No.
Name
Qualification
Yes/No
Qualifi Type of work Professional Training
in
radiation
cation carried out
training & safety ? If yes, duration
experience
year and institution
4. Persons who were contacted/
available during survey
:
II. Details of radioisotopes indented :
Radioisotope
Code
Physical form
Maximum activity Specific uses
in storage at any and quantum of
time
activity used
for operation
III. Personnel Monitoring Service (mark ‘X’ whichever is applicable)
1. All persons are enrolled
2. Partially enrolled
3. Not enrolled
If enrolled, do they wear the badges regularly while working ? : Yes/No
Are the used badges returned regularly
:
Yes/No
IV. Monitoring instruments available
Name
Make & Measurement Whether in Date of last
Model
range
working
calibration
condition ?
V. Layout of radioisotope laboratory (a rough sketch may be attached)
Number of rooms
:
Total area
:
RSD's approval reference No.
:
VI. General facilities provided
1. Separate rooms are provided for each operation such as storage, handling, counting, etc.
Yes/No
2. Storage facility is satisfactory (shielding and general safety)
Yes/No
3. Ventilation provided is satisfactory
Yes/No
4. Doors & walls are painted with washable paint
Yes/No
5. Illumination in the laboratory is satisfactory
Yes/No
6. Sinks provided are with smooth surface
Yes/No
7. Sinks are directly connected to main sewage/delay tanks
Yes/No
8. Taps to sinks are foot/elbow operated
Yes/No
9. Floor is covered with linoleum/PVC sheets
Yes/No
10. Work table surfaces are covered with smooth, non-absorbing material Yes/No
(example: Sunmica, Formica, Stainless Steel, etc.)
11. Any unwanted material/furniture/objects lying in the radioisotope
laboratory
Yes/No
VII. Radioisotope handling facilities
1. i) Fumehood with proper exhaust system is available
Yes/No
ii) Filter is provided in the fume hood exhaust system
Yes/No
2. Glove box is available
Yes/No
3. Remote handling tools available
i) Can Qpener
Yes/No
ii) Decapper
Yes/No
iii) Tongs/tweezers/forceps
Yes/No
4. Trays with smooth surface
Yes/No
5. Surgical gloves
Yes/No
6. Laboratory coats
Yes/No
7. Shoe covers/shoe barriers
Yes/No
VIII. Area monitoring and checking of contamination
1. Make & model of monitoring instrument used :
2. Exposure rate levels (mRem/h or micro Sv/h)
(i)
(ii)
(iii)
(iv)
(v)
Location :
Exposure rate :
3. Contamination levels
IX. Radioactive waste collection and disposal
1. Foot operated waste bins with disposable polythene lining inside are
(for solid waste collection)
Yes/No
2. Polythene carbouys are available (for liquid waste collection)
Yes/No
4. Adequate shielding is provided for radioactive waste storage containers
Yes/No
5. Method of disposal of radioactive waste (mark 'X' whichever is applicable)
Solid
:
i) Disposal along with other general waste
ii) Special disposal mode available (specify)
iii) Frequency of disposal : daily weekly, monthly, etc.
Liquid
:
i) Disposed into the sink daily
ii) Special delay tank facility available
iii) Stored for ……….. weeks/months and disposed into
the disposal line
iv) Any other method (specify)
5. Decontamination kit available
Yes/No
X. Records
1. Radioisotope inventory log book is maintained (isotope intended and
used, with date and activity)
Yes/No
2. Periodic checking of contamination is carried out and records available Yes/No
3. Unusual observations, if any
Yes/No
XI. Date of last survey (Comments on implementation of earlier recommendation)
XII. Overall assessment of radiation safety
XIII. Survey recommendations
XIV. Name(s) of person(s) conducting the survey
1.
2.
XV. Signature(s) of person(s) conducting the survey
1.
2.
Format of radiation protection survey and contamination check
Location
Exposure rate (mR/h)/
count rate (cpm)
Instrument Date of Checked by
used
checking
APPENDIX-3
CONTAMINATION MRASUREMENT AND DECONTAMINATION
PROCEDURES
1.
AIM : To establish surface contamination and examine the possibility of
decontaminating different surfaces.
2.
a.
REQUISITES :
GM counter
b.
Radiation counting system
c.
Metal, wood and glass surfaces
d.
Decontamination agents: water, teepol, Radiacwash, EDTA (2 % weight per volume)
and 0.1 N HCl (for metal surfaces 0.1 N HNO3)
e.
Cotton, gloves, pair of tongs
f.
Plastic tray to keep contaminated surfaces.
g.
A polythene bag inside a foot-operated dustbin to collect radioactive waste.
3.
THEORY :
Radioactive contamination results basically by contact between radioactive material and.
any surface, whenever radioisotopes in open form are handled. Direct monitoring of
surfaces, for beta ai1d gamma contamination is done where the instrument probe can be
directly held just over the contaminated area. For this purpose, a GM counter having
proper window thickness can be used at fixed geometry. For alpha contamination, a
probe with ZnS scintillator or a thin window GM counter at a distance of 5 mm from the
contaminated area can be used.
Indirect monitoring is done where direct monitoring is not feasible. For this purpose, a
swipe is taken from the contaminated area and counted using a suitable detector. It is
assumed that 10 to 20 per cent of the contamination is removed by a smear.
The experiment consists of assessing the contamination by direct method and checking
the effectiveness of various decontaminating agents for removal of contamination, due to
beta source, from different surfaces.
4.
PROCEDURE :
a.
Mark an area of to cm diameter at the centre of the given surface.
b.
Dispense 1 microcurie of phosphorous-32 and spread it uniformly within the marked
area. Place it under infrared lamp to dry.
c.
d.
Take background counts.
Monitor the contaminated surface over a fixed geometry.
e.
f.
Take a dry swipe with a filter paper (2 cm dia.) using a pair of tongs and count it.
Determine ,the fraction of the original contamination removed by the swipe.
Decontamination :
Decontamination is done step wise.
Wet the surface with water and scrub with a small amount of cotton using a pair of tongs. Dry
the surface and monitor. Repeat the procedure once more.
Repeat this procedure with Teepol(soap). Radiac wash, EDTA and 0.1 N HCI (For metal surface,
0.1 N HNO3). Wash the surface with water after each step. Repeat the counting with the same
geometry after each step and take the average of 3 to 4 readings.
After each decontamination stage, the decontamination factor F1} is calculated.
count rate before the first decontamination
F = -------------------------------------------------------count rate after the particular decontamination
Tabulate the results as in table I and calculate F1
Table 1
G.M. Counter Sr. No. : ________________ Operating voltage: ________________
Counting System Sr. No. : _____________Background(B): ______________ Counts/min
Surface
Counts before
decontamination
C0 (cpm)
Counts after each decontamination step with
Water
1
Soap
2
1
2
Radiac
1
2
EDTA
1
2
C0 - B
Decontamination Factor (FD) = ----------------------C-B
The residual contamination RR (in per cent) is calculated by the relation RR = l/FD X 100. The
decontamination factor, arrived at after the last decontamination stage is the only one used for
assessment of the sample. Assessment is made in accordance with Table 2.
Table 2
Possibility of Decontamination
very good
FD
100
100-50
Good
50-25
moderate
25
Poor
Calculation of contamination level :
It is possible to calculate the contamination level of the, surface in the following way :
a. Keep the standard phosphorous-32 source at a distance of 3 cm (same distance used for
the experiment) below the GM counter for 1 minute. Let the countrate be X counts/min.
b. Determine the background countrate. Let this be Y counts/min.
c. Repeat steps (a) and (b) at least four times and take the average count rate.
d. Let Z be the dpm of the standard source.
Efficiency of the counter in percentage is given by:
E%
(X-Y) x 100
= ---------------------Z
Contamination level A =
cpm after last decontamination
--------------------------------------- x 100 Ci
E x 2.2 x 106
e. Measure the area of contamination. Let it be B cm2
Contamination level = A/B Ci/cm2
The derived working level for alpha contamination is 10-5 Ci/cm2 and for beta 10-4 Ci/cm2.
However, it is recommended that contamination level is brought down to background level.
APPENDIX-4
CALIBRATION OF RADIATION MONITORS
1.
AIM :
To calibrate radiation monitors using standard radioactive sources.
2.
REQUISITES :
a.
b.
c.
d.
e.
Ion chamber type survey meter (Gun Monitor).
Geiger Muller survey meter.
Pocket dosimeters of various ranges.
Pocket dosimeter charger.
Radioactive sources.
3.
THEORY :
Radiation monitors are portable, battery operated and they generally measure exposure or
exposure rates. Calibration of these instruments is usually done by the manufacturer.
However, due to many factors such as changes in detector insulation, climatic effect on
components, etc. the original calibration ceases to be valid. Therefore in order to get
accurate readings, it is essential to recalibrate the instruments from time to time. The
calibration of radiation monitors consists of checking the electronic circuits by feeding
simulated pulses and checking outputs at various stages with the help of an oscilloscope.
Finally, the combined response of detector and measuring circuit is seen with the help of
standard radioactive sources, whose exposure rate at various distances can be calculated.
If the readings are not correct, then the calibration is done either by adjusting the
potentiometer specially provided for this purpose or modifying it with correction factor
(ratio of correct value to the observed reading). These values can be used later for
determining the correct exposure or exposure rate. It is also necessary to check the response of the instrument at various radiation energies expected to be encountered.
In this experiment, the following instruments, whose brief description is given, will be
calibrated. Assuming that the electronic circuit is working properly, only calibration will
be done, with standard radioactive sources.
(a)
Ion Chamber Type Survey Meter (Gun Monitor) :
This instrument consists of a 400 cc cylindrical ionization chamber detector and a d.c.
amplifier to measure the ionization current produced in the chamber when it is exposed to
radiation. This instrument can measure gamma ray exposure rates ranging from 50 mR/h
to 5000 mR/h, with an accuracy of  20 %. The instrument is energy independent
between 250 keV to 1.3 MeV for gamma rays.
(b)
Geiger Muller Survey Meter :
This is a portable area monitor consisting of a Geiger Muller (GM) counter, a
preamplifier, a pulse shaper, a countrate meter and a meter display. It has three ranges of
0.2,2 and 20 inR/h. The main advantages of this instrument are its ruggedness and high
sensitivity. The disadvantages are its energy dependence, finite dead time and paralysis at
high exposure rates.
(c)
Pocket Dosimeter :
This is a small self reading dosimeter consisting of a 2 cc condenser chamber, a quartz
fiber electrometer with a graduated scale and a pair of lenses working as a microscope. It
works on the principle of gold leaf electroscope. Initially the chamber is fully charged
and the fiber is brought to zero on the scale. When exposed to radiation, the chamber gets
discharged' partially and the fiber deflects to the right indicating the exposure on the
scale. It is energy dependent below 300 ke V and measures integrated exposure. Pocket
dosimeters having ranges of 200 mR, 300 mR, 1 R. 5 R, 10 R full scale are available.
The following formula can be used to calculate exposure rates at various distances:
KxS
E = ------------d2
where E = Exposure rate in mR/h
S = Activity of source in mCi
d = Distance in em.
K = Specific gamma ray constant for the source used in R/h/mCi at 1 cm.
Specific gamma ray constants ‘K’ for some radioisotopes are given in Table 1.
Table 1
Source
Caesium-137
Cobalt-60
Iridium-192
4.
Specific gamma ray constant
3.2
13.2
4.8
PROCEDURE :
a. Place the standard radioactive source on the axis of the cylindrical chamber of the
gun monitor at a distance do from the end of the chamber as shown in Fig. 1 (a).
b. Choose the distance ‘d’ such that the deflection in the meter in a particular range is
more then half the scale.
c. Observe the reading on the meter and compare it with the exposure rate calculated by
the above formula.
d. If the readings do not tally, adjust the calibration potentiometer, such that the meter
reads, the calculated value.
e. Check the calibration in other ranges also. This is done by placing sources of
appropriate activity and noting down the observed readings.
f. Follow the above procedure for calibration of GM survey meter [see Fig. l (b) and
pocket dosimeter [see Fig.1 (c)]. The distance‘d’ should be measured from the central
axis of the cylindrical detectors.
g. Tabulate the readings in the respective tables.
x
x = 4 cm
d1
d2
d3
(a) Ion Chamber Survey Meter (Gun Monitor)
GM detector
S1
d1
S2
d2
S3
d3
(b) GM Survey Meter
(c) Pocket Dosimeter
S1, S2 & S3 are the source position at distance d1; (d1+d2); & (d1+d2+d3)
Fig.1 : Calibration setups for Radiation Monitors.
5. OBSERVATIONS :
Table 2 : Ion Chamber Survey Meter
Sr.
Range
Distance
No.
‘d’ in cms
Reference
reading ‘A’
Observed
reading ‘B’
Calibration
factor A/B
Table 3 : GM Survey Meter
l.
Range
Distance ‘d’
No.
in cms
Reference
reading ‘A’
Observed
reading ‘B’
Calibration
factor A/B
Table 4 : Pocket Dosimeter
Sl.
Range
Distance ‘d’
No.
in cms
Reference
reading ‘A’
Observed
reading ‘B’
Calibration
factor A/B
In case of pocket dosimeter, repeat the readings with dosimeters of various ranges viz. 250 mR,
500 mR, 1 R, 5 R, and 10 R.
In order to study the energy dependence, use sources of different energies and find the
correction factor for each energy. Draw a graph of energy versus correction factor for each
instrument.