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TRANSCRIPT
Training Course
on
SAFETY ASPECTS IN RESEARCH APPLICATIONS OF
IONISING RADIATIONS
Lecture Notes
ORGANISED BY
INDIAN ASSOCIATION FORRADIATION PROTECTION
IN COLLEBORATION WITH
Radiological Physics & Advisory DivisionBhabha Atomic Research Centre
CT&CRS, Anushaktinagar,
8 . 1
Mumbai – 400 094.
8 . 2
Training Course
on
SAFETY ASPECTS IN RESEARCH APPLICATIONS OF
IONISING RADIATIONS
Lecture Notes
Compiled by P.K.Gaur
ORGANISED BY
INDIAN ASSOCIATION FORRADIATION PROTECTION
IN COLLEBORATION WITH
Radiological Physics & Advisory DivisionBhabha Atomic Research Centre
CT&CRS, Anushaktinagar,
8 . 3
Mumbai – 400 094.
8 . 4
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 . 5
1. RADIATION PHYSICS
ATOMIC STRUCTUREMatter 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.
Fig. 1.1 : Oxygen Atom
8 . 6
electrons protons
neutrons
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
with contributions entirely from protons
and neutrons. The mass of a proton and a
neutron are 1.6723 x 10-24gm and
1.67747 x 10-24gm respectively. There
are 92 elements in nature. These
were formed at the time earth was
formed. They progressively contain an
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,
: For example (Hydrogen), (Lithium)
This indicates that hydrogen atom has only one proton in its nucleus and no
neutron and lithium has three protons and three neutrons.
ISOTOPESAll 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.
Example: *radioactive isotopes
Isotopes of Hydrogen Isotopes of Cobalt (radioisotope)
Hydrogen Deuterium Tritium
Fig. 1.2: Isotopes of Hydrogen Atom.
8 . 7
p
p , n
p2n
RADIOACTIVITYThe 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 DISINTEGRATIONRadioactive 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 X-rays. Figure 1.3 shows schematically the various types of radiations
emitted during radioactive decay.
NATURE AND PROPERTIES OF NUCLEAR RADIATIONS1. 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 . 8
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 will be as follows:
e.g.
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.
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.
(-) decay when n/p ratio is high in a nucleus
8 . 9
Beta particlesAlpha particles
Gamma Rays
+ (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 will be as follows:
e.g.
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
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
8 . 10
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 DECAYThe 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
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-LIFEIn actual practice, the rate of decay of a radioisotope is usually in terms of its half-
life, 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 . 11
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 half-life, given by the relation
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 . 12
Number of half lives
Rad
ioac
tivity
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 Symbol Type of Decay & Energy in (MeV)
Half-life ALI (Bq)Ingestion Inhalation
Amercium-241 241Am (5.48), (0.060) 458 y 1 x 105 5.1 x 102
Krypton-85 85Kr - (0.68), Brem (051),
10.6 y - -
Strontium-90-Yttrium-90
90Sr- 90Y - (0.54, 2.27) 28 y 7.1 x 105 1.3 x 105
Promethium-147 147Pm - (0.23) 2.6 y 7.7 x 107 4.3 x 106
Thallium-204 204Tl - (0.77) 3.76 y 1.5 x 107 3.2 x 107
Carbon-14 14C - (0.155) 5730 y 3.4 x 107 6.2 x 107
Phosphorous-32 32P - (1.71) 14 d 8.3 x 106 6.3 x 106
Cobalt-60 60Co - (0.31), (1.17, 1.33)
5.3 y 5.9 x 106 6.9 x 105
Cobalt-57 57Co Fe-X rays (0.122) 270 d 9.5 x 107 2.1 x 107
Iodine-125 125I x-ray (0.027) 60 d 1.3 x 106 2.7 x 106
Tritium-3 3H - (0.019) 12.26 y 4.8 x 108 1.1 x 109
Caesium-137 137Cs - (0.514), (0.66) 30 y 1.5 x 106 3.0 x 106
Barium-133 133Ba (0.3, 0.082) 10 y - -Sulphur-35 35S - (0.167) 87.2 d 2.6 x 107 1.5 x 107
Chromium-51 51Cr e- (0.004), (0.005) 27.8 d 5.3 x 108 5.6 x 106
Calcium-45 45Ca - (0.257) 165 d 1.3 x 107 9.5 x 106
Sodium-24 24Na - (1.39), (1037, 2.75)
15 h 4.7 x 107 3.8 x 107
Sodium-22 22Na + (0.545), (0.511, 1.275)
2.62 y 6.3 x 106 1.0 x 107
y – years, d – days, h - hours
In the table the energy of beta particles is the maximum energy (Emax) emitted.
---------
8 . 13
2. INTERACTION OF IONISING RADIATION WITH MATTER
INTRODUCTIONRadiation 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 PARTICLESAlpha 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 . 14
raised to a higher energy state, the process is called excitation. Figure 2.1 shows
schematically the process of ionization by an alpha particle.
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 . 15
Fig 2.1: Ionisation by alpha particles
α particle
α- particle
Electron
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 – 0.106Radiative 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.
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
8 . 16
Beta Particle
NucleusBremsstrahlung
hν
Fig 2.2 : Process of bremsstrahlung
much less 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 RADIATIONSThese 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 = hwhere 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. Hence for high atomic number
8 . 17
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 . 18
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
8 . 19
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
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.
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,
8 . 20
Compton scattering
ScatteredPhoton
Photoelectron
Photoelectric Effect
Pair Production
Incident Photon
Incident Photon
Incident Photon
h 0.51 MeV
h 0.51 MeV
(a)(b)
(c)
Compton Electron
I = I0 e-x
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.
8 . 21
Incident photon Fluence
Collimator
Absorber
Scattered Photons
Transmitted Photons Fluence
Fig. 2.4: Attenuation of Electromagnetic Radiation
3. The pair production starts at 1.02 MeV and increases with energy.
4. There is a dip in the curve, which corresponds to a transition from decreasing
predominance of Compton interaction and increasing predominance of pair
production process.
--------------
8 . 22
10
1.0
0.1
Mas
s Atte
nuat
ion
Coe
ffic
ient
(cm
2 /g)
Photon Energy (MeV)
0.1 1.0 10 100
Lead
Water
Fig. 2.5: Mass Attenuation Coefficient of water and Lead for different Photon Energy
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 . 23
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 . 24
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 . 25
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 Rad. Weighting Factor
Photons all energies 1
Electrons, muons all energies 1
Neutrons energy < 10 keV
10 keV – 100 keV
100 keV – 2MeV
> 2 MeV to 20 MeV
> 20 MeV
5
10
20
10
5
Protons, other than recoil protons > 2MeV 5
Alpha particles, other than fission
fragments of heavy nuclei
All energies 20
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
Exposure R(Roentgen) C Kg-1(Coulombs/ Kg) 1 R = 2.58 X 10-4 C/Kg
Dose rad Gy (Gray) 1 Gray = 100 rads
Equiv. Dose rem Sv (Sievert) 1 Sv = 100 rems
8 . 26
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
Tissue /Organ Tissue Weighting Factor, (wT)
8 . 27
Gonads 0.20 Bone marrow 0.12 Colon 0.12 Lung 0.12 Stomach 0.12 Bladder 0.05 Breast 0.05 Liver 0.05 Oesophagus 0.05 Thyroid 0.05 Skin 0.01 Bone surface 0.01 Remainder* 0.05* Other tissue/organ which are not included in the table
------------
8 . 28
4. RADIATION DETECTORS & MONITORS
RADIATION PROTECTION INSTRUMENTSIonising 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 rateb. Dosimeters: used to measure cumulative external exposurec. 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, Geiger-
Muller Counters)
8 . 29
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 DETECTORSThis 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
8 . 30
used for 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
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
8 . 31
Display
(-ve)Outer Electrode
(+ve) Central Electrode
1.56 mGy/h
High Voltage
Electron Positive ion
Radiation
Insulator
in case of contamination monitor gives the count-rate. These instruments are used 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 DETECTORSThe 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
8 . 32
excited. The excited molecules emit photons and then return to their ground state. The 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 (tri-phenyl 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 DETECTORSThe 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
8 . 33
semiconductor as compared to number of ion pairs produced in air for same gamma
radiation energy.
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 DETECTORSPhotographic 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.
8 . 34
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
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 INSTRUMENTSRadiation 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
8 . 35
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
(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
8 . 36
sufficient activity to provide calibration on all ranges of concern. Before calibration the
instrument is checked for radioactive contamination, condition of batteries, loose or
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 . 37
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 micro-
organism, 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 . 38
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 bio-
molecules 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 . 39
Fig. 5.1 : A Typical Cell.
Fig. 5.2 Human chromosomes (Male 44 + XY)
Fig. 5.3: Human chromosome (Lymphocyte) with aberrations
(d: dicentric, Ace: acentric fragments, r : ring)
8 . 40
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
impair the function of tissue and organ.
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 . 41
BIOLOGICAL EFFECTS OF RADIATION IN MANThe 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 . 42
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 . 43
Table 5.1 : Biological Effects of Acute Exposure to Radiation
A. Whole Body Irradiation (low LET radiation)
Dose Range Immediate EffectLess than 0.1 Gy No detectable effectAbove 0.1 Gy Chromosome aberrations detectableAbove 0.5 Gy Above effect plus transient reduction in WBC
count granuclocyte count.Above 1 Gy All the above plus nausea, vomiting, diarrhea
(NVD), loss of appetite, radiation sickness: recovery probable.
3-5 Gy All the above with increased severity. Death of 50 percent of exposed population in about 60 days.
5-10 Gy Severity of above effects increases. Almost 100 percent death (at the higher dose).
B. Local Irradiation (Low LET radiation)
Dose Region Effect0.15 Gy Testes Temporary sterility3.5-6.0 Gy Testes Permanent sterility1.5-2.0 Gy Ovaries Temporary sterility2.5-6.0 Gy Ovaries Permanent sterility3 Gy Hair follicles Epilation (temporary)5 Gy Eye Cataract (after 5-10 years)6 Gy Skin Skin erythema/depilation10-20 Gy Skin 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.
8 . 44
Because of 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. .
8 . 45
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
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
ColonDose (Sv)
Subjects1950-1990
ObservedDeaths
Expected Background
Excess Deaths
0(<0.005) 36,459 3,013 3,055 -420.005 - 0.1 32,849 2,795 2,710 850.1 - 0.2 5,467 504 486 180.2 – 0.5 6,308 632 555 770.5 – 1.0 3,202 336 263 731.0 – 2.0 1,608 215 131 84> 2.0 679 83 44 39Total 86,572 7,578 7,244 334
Donald Pierce et.al (RERF update, vol 8, p.10)
Table 5.3 : Observed and Expected Deaths of Leukemia
MarrowDose (Sv)
Subjects1950-1990
ObservedDeaths
Expected Background
Excess Deaths
0(0.005) 35,458 73 64 90.005 – 0.1 32, 915 59 62 -30.1 – 0.2 5,613 11 11 00.2 – 0.5 6,342 27 12 150.5 – 1.0 3,425 23 7 16
8 . 46
1.0 – 2.0 1,914 26 4 22> 2.0 905 30 2 28Total 86,572 249 162 87
Donalt Pierce et.al (RERF update, vol.8, p.10)
Table 5.4 : Estimated Lifetime Risk of Radiation Induced Stochastic Effects in Man (Chronic Exposure)
----------
8 . 47
Effect Risk per million per mSv Spontaneous risk per million
Cancer 50 1 x 105 – 2 x 105
Hereditary 10 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
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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 PROTECTIONImplementation 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 . 49
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 cm2
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 . 50
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.
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.
Table 6.1: Dose limits
Application Annual Dose LimitOccupational Public
Effective dose 20 mSv per year, 1 mSv per year
8 . 51
Annual equivalent Dose Individual Organs Eye Lens
Skin
Hands and feet
Equivalent dose
Pregnant Women
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
150 mSv 15 mSv
500 mSv* 50 mSv
500 mSv** -
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 . 52
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, 14C). 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 . 53
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)
Annual dose limit for
Whole SingleBody organExpo.(Sv) (Sv)
Testis :
Temp. Sterility
Perm. Sterility
Ovaries: Sterility
Lens : Cataract
Bone marrow:
Depression of
Hematopoesis
Fatal aplasia
0.15
0.35
2.5 - 6.0
5.0
0.5
1.5
0.4
2.0
>0.2
>0.15
>0.4
>1
0.02 0.1
0.02 0.1
0.02 0.1
0.02 0.15
0.02 0.17
0.02 0.17
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 . 54
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 Annual Effective DosemSv
Annual Effective DosemSv
Cosmic rays Terrestrial Radiation External Outdoor Indoor Internal Inhalation Radon Thoron U,Th Ingestion K-40 U,Th etc.
0.360.41
1.26 0.07 0.01
0.18 0.06
0.070.39
Total 2.36 0.46
Fig 6.1: Percentage Contribution of Natural and Man Made Radiation to Background
--------------
8 . 55
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
8 . 56
gamma ray 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.
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 57Co? 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 = ?
Therefore
Here I1 = 2.0 mGy/h, D1 = 1 cm, D2 = 10 cm
8 . 57
Therefore, = 0.02 mGy/h
Example 2 : What is the distance required to reduce the radiation level from a 20 MBq 60Co 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
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 =
= 0.3 R/h
= 300 mR/h
Weekly permissible exposure = 40 mrad
Allowable time = 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
8 . 58
permissible exposure limit will be received in two such operations. Instead, if a 30 cm
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
8 . 59
energetic gamma radiation, the photo electric effect in high in lead (Pb). Material of
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.
8 . 60
Tenth Value Thickness (TVT): Another specific thickness which is convenient to
use 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 MONITORINGThe 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
8 . 61
to 44 mGy/h) full scale and also have audio indication. A low range in the instrument is
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 exposure-
rate 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 . 62
Table 7.1 : Radiation Monitors Commonly used in Radioisotope Laboratories
Name of the instrument
Detector Range Use Supplier*
MINIRAD GM counter(4 ranges)
0-5 R/h Routine area monitoring, Leakage
radiation level around large
sources.
Pulsecho system
MR-4500 - do - 0-50 R/h - do - ECILβ-γ Exposure Rate Meter(SM-140)
Ionization Chamber
0-5 R(3 ranges)
- do - - do -
RADMONType: RM 701
GM counter 0-2- R/h(5 ranges)
- do - Nucleonix Systems Pvt.
Ltd.Radiation Survey
Meter- do - 0-5 R/h - do - PLA Electro
App. Pvt. Ltd.Gun Monitor
(Exposure Rate Meter) GM-125
Ionization chamber
0-5R/h - do - - do -
Radiation Survey Meter PRM01215
NaI scintillation
detector
0-1 mR/h Background radiation measurement: low level radiation area
monitoring
- do -
GM Survey Meter (MR-121)
GM counter 0-20 mR/h
- do - ECIL
DIGICON Contamination
Monitor CM 710
GM counter 0-1000 cps
Beta/gamma contamination
monitoring
Nucleonix systems Pvt.
Ltd.
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 . 63
instrument.
INTERNAL HAZARD EVALUATION & CONTROLRadioisotopes 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 . 64
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 . 65
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 . 66
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 . 67
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 14C. 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 131I
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 . 68
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 131I or 125I). 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 . 69
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 non-porous 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
8 . 70
disposed off 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.
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 ContaminationLocation Beta Bq/cme Alpha Bq/cm2
Skin 1.5 1.0Hands (Total) 350 Bq 250 BqClothes1. Personal 2 0.5
8 . 71
RESPIRATORY TRACT
BLOODPULMONARYLYMPH NODES
GA
STR
OIN
TEST
INA
L TR
AC
T
EXHALATION
INGESTIONINHALATION WOUNDS
BILE LIVER KIDNEY SKIN OTHER ORGANS
FECESURINE
SWEAT
2. Plant 6 2Shoes1. Personal2. Plant
0.376
0.0372
Floor 3.7 0.37
Table 7.3 : Exposure Rate Constant for Certain Gamma Emitters
Isotope Half Life
Gamma Ray Energy (MeV) Exposure rate ConstantR/h/mCi at 1 cm
mGy/h/MBq at 1 cm
22Na 2.6 y 1.274 (100%) 12.0 2.824Na 15.0 h 2.57 (100%) 1.37 (100%) 18.4 4.442K 12.5 h 1.53 (18%) 01.4 0.3346Sc 84.0 d 1.119 (10%), 0.887 (100%) 10.9 2.6
52Mn 5.7 d 1.43 (100%, 1.44 (6%), 1.34 (6%), 0.935 (84%), 0.744 (82%), 0.51
(58%)
18.6 4.4
57Co 267 d 0.136 (9% + 1% IC), 0.122 (89% + 1% IC), 0.0144 (6% + 84% IC)
00.9 0.2
60Co 5.27 y 1.33 (100%), 1.17 (100%) 13.2 3.165Zn 245 d 1.14 (49%), 0.51 (3.4%) 02.7 0.64125I 60 d 0.035 (7% + 93 % IC) 00.7 0.17131I 8.06 d 0.36 (80% + 1% IC), 0.28 (5%), 0.08
(2% + 4 % IC), 0.72 (3%), 0.54 (9%)02.2 0.52
137Cs 30 y 0.662 (85% + 10% IC) 03.3 0.78141Ce 32.5 d 0.145 (49%) 00.35 0.083192Ir 74.0 d 0.317 (72%), 0.468 (47%), 0.308
(28%), 0.296 (26%), 0.604 (9%), 0.612 (6%), 0.589 (5%)
04.80 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
8 . 72
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 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
8 . 73
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
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 . 74
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 59Fe, 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 . 75
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 FEATURESSite
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 . 76
Table 8.1 : classification of IsotopesBasis – Relative Radio-toxicity per Unit Activity
GROUP I210Pb 210Po 223Ra 226Ra 228Ra 227Ac 227Th 228Th230Th 231Pa 230U 232U 233U 234U 237Np 238Pu239Pu 240Pu 241Pu 242Pu 241Am 243Am 242Cm 243Cm
244Cm 245Cm 246CM 249C- 250C- 252Cf
GROUP II22Na 36Cl 45Ca 46Sc 54Mn 56Co 60Co 89Sr90Sr 91Y 95Zr 106Ru 110m Ag 115mCd 114m In 124Sb
125Sb 127m Te 129m Te 125I 126I 131I 133I 134Cs137Cs 140Ba 144Ce 152Eu 154Eu 260Tb 170Tm 181Hf182Ta 192Ir 204Tl 207Bi 210Bi 211At 212Pb 224Ra228Ac 230Pa 234Th 236U 249Bk
GROUP III7Be 14C 18F 24Na 38Cl 31Si 32P 35S41A 42K 43K 47Ca 47Sc 48Sc 48Vr 51Cr
52Mn 56Mn 52Fe 55Fe 59Fe 57Co 58Co 63Ni65Ni 64Cu 65Zn 69m Zn 72Ga 73As 74As 76As
77As 75Se 82Br 85m Kr 87Kr 87Kr 85Sr 91Sr90Y 92Y 93Y 97Zr 93mNb 95Nb 99Mo 96Tc
97mTc 97Tc 99Tc 97Ru 97Ru 105Ru 105Rh 103Pd109Pd 105Ag 111Ag 109Cd 109Cd 115m In 113Sn 125Sn
122Sb 125m Te 127Te 129Te 131m Te 132Te 130I 132I134I 135I 135Xe 131Cs 136Cs 131Ba 140La 141Ce
143Ce 142Pr 143Pr 147Nd 149Nd 147Pm 149Pm 151Sm153Sm 152mEu 155Eu 153Gd 159Gd 165Dy 166Dy
166Ho 169Er 171Er 171Tm 175Yb 177Lu 181W 185W187W 183Re 186Re 188Re 185Os 191Os 193Os 190Ir194Ir 191Pt 193Pt 197Pt 196Au 198Au 199Au 197Hg
197m Hg 203Hg 200Tl 201Tl 202Tl 203Pb 206Bi 122Bi220Rn 222Rm 231Th 233Pa 239Np
GROUP IV3H 15O 37A 58m Co 59Ni 69Zn 71Ge 85Kr
85mSr 87Rb 91m Y 93Zr 97Nb 96mTc 99mTc 103Rh113m In 129I 131mXe 133Xe 134m Cs 135Cs 147Sm 187Re191m Os 193m Pt 197m Pt 232Th Th-Nat 235U 238U U-Nat
8 . 77
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 . 78
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 ofRadionuclide*
Prescribed Limits for Handling RadionuclidesType I Type II Type III
I < 185 kBq 185 kBq-185 MBq > 185 MBqII < 1.85 MBq 1.850 MBq-1.85 GBq 1.850 GBqIII & IV < 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 x 1.0 chemical preparations)
3. COMPLEX WET OPERATIONS (with risk of spills) x 0.1
4. SIMPLE DRY OPERATIONS (e.g. manipulation of powders x 0.1 and volatile radioactive compounds)
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 120-
150 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.
Fig. 8.1 : Typical Layout of Radioisotope Laboratory
10 .4
RADIOISOTOPESTORAGE
SOLID WASTESTORAGE
LIQUID WASTESTORAGE
WO
RK
TAB
LE
SINK
FUM
E H
OO
D
CHANGE ROOM
CM
DWB WB
D
D
STORAGE AREA
D
DD
10 .5
SOURCE SEORAGEAND
HANDLING
COUNTING ROOM
FUME HOOD ISOTOPESTORAGE
DD
(a)
SOURCE STORAGEAND
HANDLING
FUMEHOOD
BETACOUNTING
GAMMACOUNTING
ISOTOPESTORAGE
DD
D
(b)
Fig 8.2: Typical One room Radioisotope Laboratory
10 .6
HHOT
HOT ROOM
GAMMA COUNTING
BETA COUNTING
OFFICE
WASH
AUTO RADIOGRAPHY
GAMMA LAB
HARD BETA LAB
MASS SPECTROMETER
SOFT BETA LAB
FUME HOOD GLOVE BOX
D D
D D
D
D
D
DD
D
D
D
Fig 8.3: Typical eight rooms Radioisotope Laboratory
1. Hood Interior 2. Strippable Paint3. Service Outlets 4. Cup Sink5. Controls for Service Outlets 6. Electrical Outlets7. Cabinets 8. Filter & exhaust9. Movable Sash
Fig. 8.4 : Radiochemical Fume hood
10 .7
180 cm165 cm
90 cm
4
2
16
5
3 7
8
9
1. Glass Window 2 Service Inlets 3. Fluorescent Lamp 4. Electric Power Panel 5. Filter 6. Exhaust outlet 7. Air Lock 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
90 cm
45 cm
60 c
m
7
8
2
4
3
1
5
6
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 OFFICERThe 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 WASTESIn 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
Table 10.2 : Categories of Gaseous Wastes
10 .14
Category Activity levelA (MBqM-3)
1 A 10-10
2 10-10 < A 10-7
3 10-7 < A 14 10-3 < A 15 1 < A
Category Activity levelA (MBqM-3)
1 A 10-6
2 10-6 < A 10-2
3 10-2 < A
Table 10.3 : Categories 01 Solid Wastes
CLASSIFICATION OF WASTES AND METHODS OF DISPOSALRadioactive 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
WASTEShort-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 WasteFor the disposal of solid radioactive waste, segregation should begin at the
Category Radiation dose on the surface of wastes
D (mGy/h)
Remarks
1. D 2 x 10-3 - emitters significant emitters insignificant2. 2 x 10-3 < D 2 x 10-2
3. 2 x 10-2 D4. activity expressed in MBq/m3 -emitters dominant -
emitters insignificant
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. A i/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 131I, 3.7 MBq of 3H and 37 MBq of 32P 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
Substituting the values
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 wastesShort-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 14C
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 carcassesFor 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 foliageThe 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 Maximum activity in a pit (MBq)s
Radionuclide Maximum activity in a pit (MBq)
3H 9250 99Mo 37014C 1850 106Ru + 106Rh 37
24Na 370 124Sb 3732P 370 125I 3735S 1850 131I 37
36Cl 37 131Xe 3745Ca 370 137Cs + 137m Ba 3760Co 3700 144Ce + 144Pr 3785Kr 3700 170Tm 37059Fe 370 192Ir 37089Sr 37 210Po 3.7
90 Sr+ 90Y 3.794Zr + 95Nb 370
INCINERATIONIncineration 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, 14C) 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 Maximum limit on total discharge per day
(MBq)
Average monthly concentration of radioactivity in the discharge
(MBqM-3)3H 92.5 370014C 18.5 740
24Na 3.7 22232P 3.7 18.535S 18.5 74
36CL 0.37 7445Ca 3.7 10.160Co 0.37 37.089Sr 0.37 11.1
90Sr + 90Y 0.037 0.14894Zr + 95Nb 3.7 74
99 Mo + 99m Tc 3.7 185106Ru + 106Rh 0.37 14.8
124Sb 0.37 25.9125I 3.7 22.2131I 3.7 22.2
137Cs + 137m Ba 3.7 185140Ba + 140La 0.37 29.6
144Ce 0.37 11.1170Tm 3.7 37.0
192Ir 3.7 37.0210Po 0.037 0.74
Summary1. 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 3H and 14C) 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 PACKAGESA 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 PACKAGESThe 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 PACKAGESPackages 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 reaction59Co + 1n 60Co
The product nuclide of this type of reaction is an isotope of the target element.
b. A (n,) B C reaction 130Te + 1n 131I + - +
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 reaction35Cl + 1n 35S + 1P32S + 1n 32P + 1P
The product nuclide differs chemically from the target element.
d. A (n,) B reaction6Li + 1n 3H + 4He
In this type of reaction also, the product nuclide is, chemically different from the target
nuclide.
e. Fission ProductsWhen 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.235U (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 PROCESSINGWhen 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, Cobalt-
60, 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 125I 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 CompoundsThe 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 CompoundsRadio-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 CompoundsThe 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 Major Application
24Na 23Na (n, )24Na 15.06 h , Industry51Cr 50Cr (n, ) 51Cr 27.8 d ec, Medicine6-Co 59Co(n, ) 60Co 5.2y , Radiation
Sources131I 130Te(n, ) 131Te -131I 8d , Medicine99Mo 98Mo(n, )99Mo
235U (n, f)67h , Daughter Te-
99m is used in medicine
192Ir 191Ir(n, ) 192Ir 74.37d , Radiation Sources
14C 14N(n, p) 14C 5568y Research3H 6Li(n, ) 3H 12y Research137Cs 235U (n, f) 30.17y , Radiation
Sources11C 11B(p, n) 11C 20.4m + Medicine13N 12C(d, n) 13N 9.96m + Research15O 14N(d, n) 15O 122s + Medicine123I 124Te(p,2n) 123I
127I(p,5n) 123Xe EC 123I13h EC Medicine
62Cu 62Ni(p,n) 62Cu 9.74m EC Medicine18F 18O(p,n) 18F 110m + Medicine
---------
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 Qualification
1. Head of the department :
2. Radiation Safety Officer :
Approved by AERB : Yes/No3. Details of other radiation workers
Sr.No.
Name Qualification
Type of work carried out
Professional training & experience
Training in radiation safety ? If yes, duration year and institution
4. Persons who were contacted/ : available during survey
II. Details of radioisotopes indented :
Radioisotope Code Physical form Maximum activity in storage at any time
Specific uses and quantum of 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 & Model
Measurement range
Whether in working condition ?
Date of last calibration
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 materialYes/No (example: Sunmica, Formica, Stainless Steel, etc.)
11. Any unwanted material/furniture/objects lying in the radioisotope Yes/No laboratory
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 Yes/No (for solid waste collection)
2. Polythene carbouys are available (for liquid waste collection) Yes/No
4. Adequate shielding is provided for radioactive waste storage containersYes/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 used
Date of checking
Checked by
APPENDIX-3
CONTAMINATION MRASUREMENT AND DECONTAMINATION PROCEDURES
1. AIM : To establish surface contamination and examine the possibility of decontaminating different surfaces.
2. REQUISITES :a. 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. Take background counts.d. Monitor the contaminated surface over a fixed geometry.
e. Take a dry swipe with a filter paper (2 cm dia.) using a pair of tongs and count it.f. 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 decontaminationF = --------------------------------------------------------
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 Soap Radiac EDTA1 2 1 2 1 2 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
FD Possibility of Decontamination100 very good
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:
(X-Y) x 100E % = ----------------------
Z
cpm after last decontamination Contamination level A = --------------------------------------- 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. Ion chamber type survey meter (Gun Monitor).b. Geiger Muller survey meter.c. Pocket dosimeters of various ranges.d. Pocket dosimeter charger.e. 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 re-sponse 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:
K x SE = -------------
d2
where E = Exposure rate in mR/hS = Activity of source in mCid = 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 Specific gamma ray constantCaesium-137 3.2
Cobalt-60 13.2Iridium-192 4.8
4. 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.
Fig.1 : Calibration setups for Radiation Monitors.
x
x = 4 cm d1 d2 d3
(a) Ion Chamber Survey Meter (Gun Monitor)
d1 d2 d3
S1 S2 S3GM detector
(c) Pocket Dosimeter
S1, S2 & S3 are the source position at distance d1; (d1+d2); & (d1+d2+d3)
(b) GM Survey Meter
5. OBSERVATIONS :
Table 2 : Ion Chamber Survey Meter Sr. No.
Range Distance ‘d’ in cms
Reference reading ‘A’
Observed reading ‘B’
Calibration factor A/B
Table 3 : GM Survey Meter l. No.
Range Distance ‘d’ in cms
Reference reading ‘A’
Observed reading ‘B’
Calibration factor A/B
Table 4 : Pocket DosimeterSl. No.
Range Distance ‘d’ 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.