Radiation protection in radiation oncology

Radiotherapy essential

2. INTRODUCTION
Discovery of radiation led to dramatic
advancements in medical diagnosis & treatment.
Ignorance of hazards of radiation resulted in
numerous injuries to patients, physicians and
scientists.
The growing evidence about dangers of radiation
led to efforts to guard against needless or excessive
exposure & need for setting safety standards.
The Safety Standards help to make value
judgments regarding

relative importance of risks of different kinds

arriving at a suitable balance b/w prevailing risks
& benefits

3. REGULATORY AUTHORITY
UNSCEAR (United Nations Scientific Committee on the Effects of Atomic
Radiation, 1955) established to estimate potential health risks from radioactive
fallout from atmospheric nuclear weapon tests.
ICRP (International commission on radiation protection) provides recommendations
IAEA (International Atomic Energy Agency) establishes standards of safety and
provides for the application of the standards
In India regulatory authority is AERB

5. RADIATION PROTECTION QUANTITIES
The absorbed dose is the basic physical dosimetry quantity, but is
not entirely satisfactory for radiation protection purposes.
 To account for biological effects of different radiation types on
different body tissues special radiation protection quantities used
are:

Organ dose

Equivalent dose

Effective dose

Committed dose

Collective dose

6. ORGAN DOSE
Organ dose D
T is defined as mean dose in a specified tissue or
organ T of the human body, given by:
m
T is the mass of the organ or tissue under consideration
ε
T is the total energy imparted by radiation to that tissue or organ.
T
T
m
mT
T
m
dD
m
D
T


1

7. EQUIVALENT DOSE (H
T,R)
The equivalent dose, H
T,R, in a tissue or organ T due to radiation R, is
defined as product of average absorbed dose in tissue & radiation
weighting factor
 H
T,R = w
R.D
T,R

where
D
T,R
is absorbed dose delivered by radiation R averaged over a tissue or organ T.
w
R is the radiation weighting factor for radiation type R.
For more than one type of radiation, equivalent dose is given by summing
contribution from different types of radiation :
Unit of equivalent dose is Sievert (Sv)

R
RTRRT DWH
,, .

8. EQUIVALENT DOSE (H
T,R)
The equivalent dose replaces
quantity dose equivalent.

Dose equivalent is the dose
to a point in organ
multiplied by radiation
weighting factor (w
R
).

9. EFFECTIVE DOSE
Different body tissues respond differently to radiation
Hence probability of stochastic effects from a given equivalent dose depend
upon the particular tissue irradiated.
The effective dose E is defined as the summation of tissue equivalent doses
H
T, each multiplied by appropriate tissue weighting factor W
T,
to indicate
combination of different doses to several different tissues in a way that
correlates well with all stochastic effects combined.
Unit - Sievert (Sv)
Previously the concept of Effective dose equivalent was used which defined
the effective dose to a point (as in Dose Equivalent)

TTHwE

11. COMMITED DOSE

When radio nuclides are taken into body, the resulting
dose is received throughout period of time during which
they remain in body.
The total dose delivered during this period of time is
referred to as committed dose and is calculated as a
specified time integral of the rate of receipt of the dose.

12. COLLECTIVE DOSE

Collective dose relates to exposed populations or groups
It is defined as the summation of products of mean dose
in various groups of exposed people & number of
individuals in each group.
The unit of the collective dose is the man-sievert.

Collective dose is regarded as a useful tool, in the
comparative sense, for optimization process.

14. RADIATION EFFECTS
Stochastic effects Non stochastic effects
No threshold dose. definite threshold dose
Probability rather than severity of effect is proportional to
dose.
Severity rather than probability of effect is proportional to
dose
There is probability of effect occurring even at low dosesThese are produced by relatively high doses
Generally occurs due to small modification in a single cellSeen when large number of cells are involved/ killed
Effects manifest after some time i.e. latent period b/w time
of exposure & appearance of effect.
Time course of effect manifestation depends on kinetics of
cell division in different organs.
These are somatic & hereditary effects These are somatic effects

15. RISK ASSESSMENTS: DATASETS
Dose limits are designed taking into consideration
stochastic effects i.e. radiation induced carcinogenesis &
radiation induced hereditary effects.
First risk estimates are calculated for stochastic effects
for various groups & then dose limits are designed
people exposed from the atomic bomb explosions
people exposed during nuclear and other radiation accidents
patients exposed for medical reasons
people exposed to natural radiation
workers in radiation industries

16. CANCER INDUCTION: RISK
Cancer is most important stochastic effect & often fatal.
The risk of getting cancer from radiation depends on many factors,
such as dose & how it is administered over time; the site &
particular type of cancer; & a person’s age, sex, and genetic
background.
Based on reports of UNSCEAR & BEIR V committees, the
ICRP suggests a risk estimate of excess cancer mortality
PopulationHigh dose/high dose rateLow dose/low dose rate
working 8 x 10
-2
/Sv 4 x 10
-2
/Sv
General 10 x 10
-2
/Sv 5 x 10
-2
/Sv

17. HEREDITARY EFFECTS
Ionizing radiations can produce mutations which give rise to harmful effects in future
generations.
Although mutations arise in without apparent cause, natural radiations & other
environmental agents may cause them & contribute to prevailing occurrence of
hereditary diseases.
One can’t make out whether hereditary effects are attributable to exposure from
natural or artificial radiation.
Studies of offspring of atomic bomb survivors have failed to show increase of statistical
significance in hereditary effects.
ICRP has estimated risk of genetic disorders in future generations based on animal
experiments.

18. HEREDITARY EFFECTS
Radiation increases the incidence of mutations that occur
spontaneously.
A useful, quantitative benchmark for characterizing
radiation-induced mutation rates is the doubling dose.

It is defined as the amount of radiation that produces in
a generation as many mutations as arise spontaneously.

19. WHICH IS MORE IMPORTANT
For many years (upto 1950s), the genetic effects of
radiation were considered to pose the greatest danger for
human populations exposed to low levels of radiation.

Today, the major concern is cancer

20. DETRIMENT
Radiation detriment is a concept used to quantify the harmful
effects of radiation exposure in different parts of the body.
It is determined from nominal risk coefficients, taking into
account the severity of the disease in terms of lethality and years
of life lost.
Total detriment is the sum of the detriment for each part of the
body (tissues and/or organs).
ICRP takes into account
fatal cancer
non-fatal cancer
severe hereditary effects
number of years of life lost .

21. ICRP RISK ESTIMATES

22. REFERENCE MAN
Reference Man, the ICRP model for dose calculations from the intake of
radio nuclides.
ICRP-23 (1975) concentrated on characteristics for a standard man, whereas
ICRP-89(2002) presents data for males and females of six different ages.
Reference man is defined as being between 20-30 years of age, weighing 73
kg, 170 cm in height, and lives in a climate with an average temperature of
from 10-20°C. He is a Caucasian and is a Western European or North
American in habitat and custom.”
Relatively few individuals in any group will have all characteristics close to
the reference values.
Most countries have modified the concept of reference man – e.g. Indian
Reference Man

23. REFERENCE MAN: IMPORTANCE
The concept of a ‘reference human’ help to manage many different
situations in which human beings would or could be exposed to ionising
radiations.
It enables base-line calculations of organ doses to be made for a
radionuclide incorporated under a set of very specific, well defined
assumptions.
The formalism furnishes the basis for analysis in routine monitoring and
bioassay programs throughout the world.
When applied to special situations, appropriate adjustments of some of
the assumptions can be made in order to obtain more realistic internal-
dose estimates for a particular individual.

25. DEFINITION
Radiation protection is a tool for management of measures to
protect health against the detriment (for people &
environment) generated by the use of ionizing radiation.

26. AIM
The specific objectives of radiation protection are:

To prevent the occurrence of clinically significant radiation-
induced deterministic effects by adhering to dose limits that are
below the apparent threshold levels

To limit the risk of stochastic effects, cancer and genetic
effects, to a reasonable level.

27. PRINCIPLES OF RADIATION PROTECTION
Justification
Optimisation
Limitation

28. JUSTIFICATION
“No practice involving exposure to radiation should be
adopted unless it produces sufficient benefit to exposed
individual or to society to offset radiation detriment it
causes or could cause.” i.e. practice should be justified.
Practice must result in a net benefit

29. OPTIMIZATION
Radiation sources and installations should be provided
with the best available protection and safety measures
under the prevailing circumstances, so that

the magnitudes and likelihood of exposures and

the numbers of individuals exposed
Is as low as reasonably achievable (ALARA), economic
& social factors being taken into account
protection and safety should be optimized

30. LIMITATION
The exposure of individuals from all relevant practices
should be subject to dose limits, or to some control in the
case of potential exposures

Dose limits are aimed at ensuring that no individual is
exposed to radiation risks that are judged to be
unacceptable in any normal circumstances.

31. TYPES OF EXPOSURES
For the purposes of radiation protection, ionizing radiation
exposures are divided into three types:

Medical exposure, which is mainly exposure of patients as part of
their diagnosis or treatment

Occupational exposure, which is exposure of workers incurred in
the course of their work

Public exposure - Exposure incurred by members of public from
radiation sources

33. DEFINITION

A dose limit is defined in the BSS as “The value of the
effective dose or the equivalent dose to individuals from
controlled practices that shall not be exceeded.”

34. HISTORY
In 1902, six years after discovery of x rays, first dose limit
of about 10 rad per day was recommended as this was the
lowest value that could readily be measured by fogging of a
photographic plate
September 1924 –concept of a “tolerance” dose rate for
radiation workers was introduced, a dose rate that was
considered to be one that could be tolerated indefinitely.

U.S. Advisory Committee on X-ray and Radium Protection
recommended limit on dose rate as 0.1 roentgen / day

35. DESIGN OF LIMITS: PRINCIPLE

1977 – ICRP adopted a more formal risk-based approach for setting
standards.

The approach was based on the premise that the average incremental
risk of death from radiation exposure to workers be no larger than
that from injuries to workers in “safe” industries.

The annual rate of fatal accidents in “safe” industries vary from
about 0.2×10
–4
to 5×10
–4

For an annual dose limit of 20mSv/yr or 20x10
-3
Sv/yr.

Using the total probability coefficient for workers one finds for the
average total detriment incurred by a worker i.e.
(20×10
–3
Sv y
–1
) (5.6×10
–2
Sv
–1
) = 1.2×10
–4
y
–1

This level is in the range of average annual risk for accidental death
for all industries.

36. MAXIMUM PERMISSIBLE DOSE
As defined by NCRP the Maximum Permissible Dose (MPD) is that dose
which, in the light of present knowledge is not expected to cause detectable
bodily injury to the person at any point during his lifetime.
The maximum permissible levels are not to be considered as “acceptable,” but
instead, they represent the levels that should not be exceeded.
Advantages:
Explicit acknowledgement that doses below MPD have a risk of detrimental
effects.
Acknowledged danger due to stochastic effects of radiation.
Introduced the concept of acceptable risk – probability of the radiation induced
injury was to be kept low to be easily acceptable to the individual
Allowed different levels for radiation workers and public
Allowed modifications in the advent of new knowledge.

37. INITIAL MPD

Allowed a maximum annual exposure limit of 5 rem
/year.
age-proportion formula to calculate the MPD
Accumulated MPD = 5 (Age in years - 18)
The new guidance is that the numerical value of the
individual workers’ lifetime effective dose equivalent in
tens of mSv (rem) does not exceed value of his or her age
in years.

38. The effective dose limit ensures avoidance of deterministic effects
in all body tissues & organs except for lens of eye which makes
negligible contribution to effective dose & skin which may be
subjected to localized exposures.

39. PREGNANT WORKER
A pregnant female worker should notify the employer in order that
her working conditions may be modified if necessary.
The employer shall adapt the working conditions so as to ensure
that the embryo or foetus is afforded same level of protection as
required for members of public.

For example, some institutions have developed a policy of not
assigning pregnant technologists to work with cobalt-60
teletherapy units (because of constant radiation leakage from
source housing)
from declaration of pregnancy a supplementary equivalent dose
limit of 2 mSv is applied to surface of woman’s abdomen

40. ARLI
NCRP(116) introduced annual reference level of intake (ARLI).
It is defined as “the activity of a radionuclide that, taken into the
body during a year, would provide a committed effective dose to a
person, represented by Reference Man, equal to 20 mSv.
The ARLI is expressed in Becquerel (Bq).”
Prior to ICRP 60 and NCRP (116), the term ALI was used by both
organizations. & was based on 50mSv committed effective dose
equivalent
The NCRP defines the derived reference air concentration (DRAC)
as “that concentration of a radionuclide which, if breathed by
Reference Man, inspiring 0.02m
3
/ min for a working year, would
result in an intake of 1 ARLI.”

41. METHOD TO REDUCE RADIATION
Practical control measures that can be implemented at the
workplace to control external hazards include:

Time

Distance

Shielding

42. Time
The dose accumulated by a person is directly proportional
to the amount of time spend in the radiation area.
The less time spent in a radiation environment the smaller
is the radiation dose.

Plan the work to avoid unnecessary exposure.
If necessary, a dose rate measurement or estimate can be
made and a time limitation set for the work undertaken.

43. DISTANCE
The greater the distance from a source of radiation the
smaller is the radiation dose.
For distance, the inverse square law applies, ie. for an
isotropic point source of radiation the dose rate at a given
distance from the source is inversely proportional to the
square of the distance.
Thus doubling distance from a source, decreases dose rate
by a factor of four.

44. SHIELDING
Shielding is practice of placing an attenuating medium b/w source of
ionizing radiation and person.
The attenuating medium, or shield, then minimizes the radiation
The type and amount of shielding required depends on the type and
energy of radiation emitted and its intensity.

Dense (high atomic number) materials (e.g. lead and depleted uranium) make
the most effective shields for highly penetrating radiation such as gamma
radiation.

For lesser penetrating radiation such as beta particles low atomic number
materials can be used (e.g. Perspex or aluminum).

46. LOCATION
Radiotherapy departments are usually located on the
periphery of the hospital complex

to avoid radiation protection problems arising from therapy rooms
being adjacent to high occupancy areas.
Ideally should be at the ground level

Wherever possible the treatment bunker should be
surrounded with rooms that have low occupancy.

47. ROOM SIZE
The machine manufacturer’s pre-installation manual
should provide the minimum room dimensions (length,
width and height).
Room should be large enough to allow full extension of
the couch in any direction, with room for operator to
walk around it.

48. CLASSIFICATION OF AREAS
There areas in radiotherapy are classified by BSS as :
Controlled areas - are
Teletherapy & Brachytherapy treatment rooms
source storage & preparation rooms.
In addition, these areas will require special access restrictions by
means of door interlocks and signs.
Supervised areas - are
Operating consoles of Teletherapy & Brachytherapy units &
 area where calculated exposure rates through shielding barriers are likely to
result in exposures of 1 mSv in a year
For protection calculations, the dose equivalent limit is assumed to be
0.1rem/wk for controlled areas & 0.01rem/wk for non-controlled areas.

49. TYPES OF RADIATION
Protection is required against three types of radiation:

Primary radiation- Radiation beam directly emitted from treatment
machine through collimator opening in case of external sources and
from radioactive source in case of brachytherapy
Scattered radiation - Radiation produced by scattering of primary
beam from various media such as patient, collimators, beam shaping
accessories & air.
Leakage radiation - Radiation that escapes through shielded head of
therapy unit (for accelerators leakage radiation only exists while
beam is on; for cobalt units leakage radiation is always present).

it has same intensity as that of primary radiation

50. TYPES OF BARRIERS
Primary barrier - The wall Where primary beam
strikes is primary barrier.
Secondary barrier - Protect against scattered &
leakage radiation.
Maze - Maze is a restricted access passageway
leading to the room incorporated to reduce the
radiation dose near the entrance
It ensures that photon radiation exits room
after scattering has attenuated it.
Adv. Of maze
reduces radiation dose near entrance
reduces the need for a heavy shielding door.

51. DESIGN FACTORS
Factors required for calculation of barrier thickness are :

Workload

Use factor

Occupancy factor

52. WORKLOAD

The term workload (W) is used to provide some indication
of the radiation output per week of external beam X ray
and gamma ray sources.
This is estimated by multiplying the no. of pt. with the dose
delivered at 1m.

Workload is expressed in rad/wk.

53. USE FACTOR
A use factor (U) describes the fraction of operating time during
which different beam orientations used for treatment are directed
at a particular barrier
Use factors depends on the particular use of the facility and also
on the energy used.
The following primary beam use factors are usually assumed for
external beam machines:
U (floor) = 1
U (walls) = 0.25
U (ceiling) = 0.25
For all secondary barriers U is always equal to 1, since secondary
radiation is always present when beam is on.

54. OCCUPANCY FACTOR
The occupancy factor (T) relates to the amount of time
the rooms adjacent to the treatment room or area of
interest are occupied by individuals.
Typical values are
T (offices) = 1;
T (corridors) = 0.25;
T (waiting rooms) = 0.125.

55. MATERIAL USED
Shielding should be designed by a qualified expert to ensure that
the required degree of radiation protection is achieved.
The usual materials for radiation shielding are (normal or high
density) concrete, steel, or lead.
Concrete is usually the cheapest material as it is easier to bring to
the site and use for construction.
An on-site concrete testing should be used. In new construction,
standard concrete of density 2350 Kgm
–3
should be used
If there are space restrictions, then it may be necessary to use
higher density materials such as steel or lead

58. THICKNESS OF PRIMARY BARRIER

Expression to determine the attenuation required by the barrier :
where

P is the allowed dose per week (Sv/week) outside the barrier

d is the distance from the isocentre to the outside of the barrier, in m

SAD is the source–axis distance, in m

W is the workload, in Gy/week at 1 m

U is the use factor or fraction of time beam is likely to be incident on the barrier

T is occupancy factor
The number of TVLs required to produce this attenuation is determined from:
WUT
SADdP
B
2
)(








B
LogofTVLsNo
1
.
10

59. SCATTRED RADIATION

The required barrier transmission (B
p) needed to shield against radiation
scattered by the patient is given
Where

P is the design dose limit

W is workload

T is occupancy factor

d is the distance from the radiation source to the patient, in m.

d’ is the distance from the patient to the point of interest, in m.

α is the scatter fraction defined at d. The scatter primary ratio is dependent on the
energy of the X ray beam and the scattering angle.

F is the field area incident on the patient, in cm
2
.
 400/
22
FWT
dPd
B
p




60. LEAKAGE RADIATION
The required attenuation (B
L) to shield against leakage radiation is
as follows :
P is the design dose limit;
ds is distance from isocentre to point of interest in m;
W is workload;
T is occupancy factor.
WT
Pd
B
S
L
2
1000


61. ADDITONAL CONSIDERATIONS
For external beam therapy units extent of primary barrier will be
determined by divergence of the primary beam to outside of barrier.
The primary barrier is then extended further by 30 cm on each side to
allow for small angle scatter (also termed the plume effect).
The broadening of a radiation beam beyond geometrical divergence due to the
accumulation of lateral scattering with depth
Since leakage radiation and scatter radiation are of different
energies, secondary barrier requirements of each are calculated
separately & compared to arrive at final secondary barrier thickness:
If thickness of required barrier is about same for each secondary component,
one HVL is added to the larger of the two barrier thicknesses.
If two barrier thicknesses differ by one TVL or more, larger barrier thickness is
used

62. NEUTRON PROTECTION
Considered for LINACs operating above 10 MV.
Concrete has high water content – TVL for photo neutrons half that of the
photons – additional shielding not needed.
Neutron capture produces secondary photons (capture photons):

Average capture photon energy is 3.6 MeV

Max energy is 8.0 MeV
Doors of linacs may require shielding against x-rays & neutrons scattered through
maze
Neutrons are thermalized & absorbed with a layer of about 12cm of borated
polyethylene in the door, which is followed by 2.5cm of lead to absorb gamma rays
produced by neutron capture reactions in boron nuclei.

63. TELETHERAPY ROOM: DESIGN

The treatment room door should have

a ‘fail safe interlock’ to switch off radiation
beam (i e return source to shielded position)
if door is opened during treatment
Restart of irradiation should require
both closing of door & activation of
a switch at control console

a sign which indicates the room is a
radiation area and/or contains radioactive
materials

a visible light at the door which shows if the
source is on. red for source on & green for
source off

There should be a battery operated
scatter radiation detector inside the
room which shows when the source is on

64. Viewing system either of

Mirror & glass door
arrangement

Lead glass

Two CCTV - Two cameras
are recommended – one
15° off & another above
the gantry rotation axis
for optimum patient
viewing.

65. HDR ROOM: DESIGN
In HDR brachytherapy treatment rooms, all walls are primary
barriers, since:
Source can be positioned anywhere in the room.
Radiation is emitted isotropically and is uncollimated from the source.
Primary barrier transmission factor B for an HDR brachytherapy
machine is calculated similarly to the external beam therapy except
that use factor U = 1.
P is the design effective dose.
d is the distance from the source to the point of interest.
W is the brachytherapy workload in Gy.m
2
/week
T is the occupancy factor
WT
Pd
B
2


66. 
The room should be designed so that:

Door interlock which will cause the source
to be retracted into its shielded housing if the
door is opened during the time the source is
on,

An indicator at room door & at treatment
console of source ‘ on-off’
status.

A battery operated scatter radiation detector
inside the room which shows when the
source is on

The room door should be marked for the
radioactive materials which are within &
there should be an indication of how to
contact the responsible radiation safety
individual in the event of an emergency.

68. MONITORING
Radiation monitoring is one of the most important functions for
detection and measurement of radiation and/or radioactive
contamination
Monitoring is of two types

Area monitoring -The instruments used for measuring radiation levels are
referred to as area survey meters (or area monitors)

Dose rate meters are used for this purpose. That are capable of giving direct
reading of dose equivalent rate.

Individual monitoring – the instruments used for recording the equivalent
doses received by individuals working with radiation are referred to as
personal dosimeters (or individual dosimeters).
Dosimeters measures the cumulative energy absorbed as a consequence of
exposure to ionizing radiation.

69. AIM
Radiation monitoring is carried out:

To assess workplace conditions and individual exposures;

To ensure acceptably safe and satisfactory radiological
conditions in the workplace;

To keep records of monitoring, over a long period of time, for
the purposes of regulation or good practice.
All instruments must be calibrated in terms of
appropriate quantities used in radiation protection.

70. AREA MONITORING
Area monitoring is done with portable instruments, called “survey meters”
They are primarily used to detect contamination and to determine radiation field
intensities.
Many radiation safety decisions - such as the need for decontamination, shielding,
personnel monitoring, change in work procedures, etc. - are based on radiological
survey results using survey meters.
Instruments used as survey monitors are either
Gas filled detectors
Ionization chambers
Proportional counters
Gieger-muller counters
Solid state detectors
Scintillators
Semiconductor detectors

71. DOSE RATE METER:BASIC COMPONENTS
The key components dose rate meter are:
The detector - contains a medium which absorbs
radiation energy and converts it into a signal.
Electrical charge usually forms the signal.
The amplifier - The signals from a detector may
need to be electronically amplified.
The processor - According to the type of
instrument, the processor may be a device to
measure the size or number of signals produced by
the detector. It may also translate the quantity
measured into appropriate radiological units.
The display. - The measurement is presented either
in a digital format or as an analogue display
showing a pointer on a graduated scale.
Display
processor
amplifier
Detector

72. GAS FILLED DETECTORS
A gas-filled detector consists of a
volume of gas b/w two electrodes,
with an electrical potential difference
(voltage) applied b/w the electrodes.
Ionizing radiation produces ion pairs
in the gas.
Positive ions are attracted to
negative electrode (cathode); negative
ions are attracted to positive
electrode (anode)
In most detectors, cathode is the wall
of the container that holds the gas
and anode is a wire inside the
container

73. IONIZATION CHAMBERS
Designed in many shapes & sizes & with different gas fillings.
The electrode material used, type of gas filled & its pressure &
size of ionization chamber etc. depends on radiation intensity to
be measured.
Air-filled ion chambers are used in portable survey meters, for
performing QA testing of diagnostic and therapeutic x-ray
machines.
These practical ionization chambers have

Beta window made of thin foil (3–7 mg cm
–2
).

Protective buildup cap (200–300 mg cm
–3
) made of toughened plastic or
aluminium – to improve detection efficiency for high energy photon
radiation.

74. Proportional counters- contain a gas with specific properties
Proportional counters are more sensitive than ionization chambers
hence are suitable for measurements in low intensity radiation
fields.
Commonly used in standards laboratories, health physics
laboratories, & for physics research
Seldom used in medical centers
G.M. counters -
GM survey meters are used at very low radiation levels.
In general, GM survey meters are inefficient detectors of x-rays
and gamma rays

75. INDIVIDUAL MONITORING

Individual monitoring is the measurement of the radiation
doses received by individuals working with radiation.

Used for Individuals who regularly work in controlled areas or
those who work full time in supervised areas.

Used to verify the effectiveness of radiation control practices in
the workplace.

Useful for detecting changes in radiation levels in the workplace
and

To provide information in the event of accidental exposures.

76. INDIVIDUAL MONITORING CONTD.

Individual monitoring devices are of two types:

Indirect monitoring devices - includes film badges & TLD
badges
The personal monitors are usually worn at the collar, chest or belt level.
Wearing personal monitors at belt level effectively measures radiation
dose received by trunk, but badge should not be shielded by a bench or
table when working with radiation.

Direct monitoring devices - fall into two categories:
Self reading pocket dosimeters
Electronic personal dosimeters (EPDs)

77. FILM BADGE

A film badge dosimeter consists of a photographic film & filters in a
holder.

The film usually has two emulsions of ‘fast’ and ‘slow’ sensitivities
extending the dose response from 100 mSv to 10 Sv.

The emulsions may be on the same or separate bases and sealed in paper
to prevent their exposure to light.

An identification mark printed on the wrapper appears on the
developed film.

Photographic emulsion and tissue do not absorb radiation energy in the
same proportion: film is not ‘tissue equivalent’ and must be used with a
holder.

The holder creates a distinctive pattern on the film indicating the type
and energy of radiation to which it was exposed (discrimination).

79. ADV. OF FILM BADGE

It is a permanent record. If a suspicious reading is noted,
the film may be reread.
The pattern on the film may indicate the angle of exposure,
whether the exposure was a series or single event, and the
energy, which may be correlated with the employee’s
working environment.
Separate contribution of different types of radiation is
assessed by comparing

the optical density behind a suitable filter that absorbs them &

the density through a neighboring “open window.”

80. DISADV. OF FILM BADGE
Films are also adversely affected by

light (if the wrapper is damaged)

heat

Liquids

humidity
The latent image on undeveloped film fades with time, limiting possible wearing
period
 It is not reusable.
Radioactive contamination produces non-uniform black patches on the developed film.

Due to the energy dependence, the film must be properly placed between the filters in
the badge holder or erroneously high readings will be observed.

The size of the film limits the ability to monitor fingers or eyes.

81. TLD BADGE

Thermo luminescence is a physical characteristic of certain
crystalline materials called phosphors. They absorb energy from
ionizing radiation and release it as light when heated above 100 to
200
o
C.

The intensity of the light may be measured and related to the
radiation dose of the phosphor.

TLD badge consists of a metallic card having three discs of
CaSo
4
:Dy in Teflon matrix (0.8mm thick & 13.3mm dia.)

Like film, they require filters to match their energy response to that
of tissue..

Plastic holder has three well defined areas over TLD discs

Open window

1.5mm thick plastc filter.

1mmAl +1mmCu combined filter.

82. TLD READER
Power Supply
PMT
Amplifier
Filter
Heated Cup
TL material
To High
Voltage
To ground
Recorder

83. ADV. OF TLD

Advantages (as compared to film dosimeter badges) includes:

TLDs are less affected than film badges by fading & ambient conditions
(temperature & humidity)

Able to measure a greater range of doses

Doses may be easily obtained

Quicker turnaround time for readout

Reusable

Small size

Low cost

limited energy dependence

84. QUARTZ FIBRE ELECTROMETERS

A quartz fibre electrometer provides direct reading of
cumulative exposure.

Consists of an ionization chamber that acts as a
capacitor.

Electrometer is charged prior to use.

When plugged into a charger, electrical charge flows up
the charging pin to the quartz fibre and repellor.

A light illuminates the inside of the QFE so that the
position of the quartz fibre is seen as the repellor & fibre
repel each other.

When capacitor is fully charged then the fiber is set
against zero on the scaled greticule.

When ionizing radiation ionizes the air in the chamber,
the charge on the fibre & repellor is reduced allowing the
fibre to move towards the repellor.

If the QFE is held up to light and viewed, the fibre
appears to indicate the ‘dose received’ on the reticule.

QFEs with maximum ranges of 2 mSv to 10 Sv are
available.

85. QUARTZ FIBRE ELECTROMETERS
Used to evaluate possible radiation exposure quickly.
Used in operations which involve high exposure to
ionizing radiations e.g. source change in telecobalt or
brachytherapy units.
This type of dosimeter is subject to erroneous readings due
to electrical leakage or being hit/dropped. Also, they are
not accurate for extended exposures.

86. EPD
Useful for the short term monitoring of an individual.
 These instruments have a visual readout of the
accumulated dose as well as an audible alarm if the dose
rate is too high or the cumulated dose is too great.

However, these units are expensive and easily damaged.

88. From prescription to delivery of radiation dose, there is a
combination of several manual to sophisticated computer
assisted techniques and high technology equipment
Potential for accidents in radiotherapy is very high due to
complexity in the process and equipment used.

89. HOW TO AVOID AN ACCIDENT

The equipment should be handled by authorized, qualified & trained technologists
only, and they must be aware of emergency procedures.

A clear understanding of radiotherapy equipment and its source movement system is
required to handle these emergency situations

The technologists should be alert during the delivery of the treatment. So Keep eye
and mind open.

Ensure that patient setup is precise to deliver the prescribed dose to the prescribed
treatment area and to proper patient.

Technologists should use personnel monitoring system

Make sure that the following materials are available in the area outside the
treatment room.

‘T’bar for the retrieval of the source to safe position in case of telecobalt unit.

User manual of the equipment.

Long forceps for the retrieval of brachytherapy source.

A good working Survey Meter

90. SOURCE STRUCK SITUATION

91. 
Emergency procedures for
safely removing the
source from the patient
and quickly storing it in
a safe location in the
event that it does not
retract all the way into
its source housing when
expected.

This requires that a wire
cutter to cut the source
cable and a shielded
storage container be
located inside the
treatment room,
Golden Hand Crank to
retrieve the source
Emergency button at the
Control Console

92. CONCLUSION

Radiation protection should be given due importance.
Every dose is harmful. There is no safe dose.
Ionizing radiations need to be handled with care rather
than fear.
The procedures & guidelines available to control
exposures to ionizing radiations are sufficient, if used
properly.