1. KOUSTAV MAZUMDER
MD PGT, Dept of RADIOTHERAPY
MEDICAL COLLEGE &
HOSPITAL, KOLKATA

2. INTRODUCTION
•The protection of people and the environment from the harmful effects of
ionizing radiation, which includes both particle radiation and high energy
electromagnetic radiation
•At the Second International Congress of Radiology in Stockholm in
1928, member of variuos countries were invited to send representatives to
prepare x-ray protection recommendations
•World War II was remodeled into two commissions that survive to this day:
The International Commission on Radiological Protection (ICRP)
The International Commission on Radiation Units and Measurements
(ICRU)
•U.S. Advisory Committee on X-Ray and Radium Protection (ACXRP) ,1929
which was renamed National Council on Radiation Protection and
Measurements (NCRP), on 1964

3. DEFINITIONS
Absorbed Dose:
Energy per unit Dose
Equivalent Dose:
Absorbed dose × radiation weighting factor
A radiation weighting factor (WR) is a dimensionless multiplier used to place
biologic effects (risks) from exposure to different types of radiation on a common
scale

4. CONTD……
Effective Dose
Sum of equivalent doses to organs and tissues exposed, each multiplied
by the appropriate tissue weighting factor (WT)
Tissue Weighting Factor (W T), which represents the relative contribution of each
tissue or organ to the total detriment resulting from uniform irradiation

5. CONTD…..
Committed Equivalent Dose
ICRP defined the committed equivalent dose as the integral over 50
years of the equivalent dose in a given tissue after intake of a radionuclide
Committed Effective Dose
If the committed equivalent doses to individual organs or tissues
resulting from the intake of a radionuclide are multiplied by the appropriate
tissue weighting factors and then summed, the result is the committed
effective dose.
Collective Equivalent Dose
Product of the average equivalent dose to a population and the
number of persons exposed
Collective Committed Effective Dose
The integral of the effective dose over the entire population out to a
period of 50 years is called the collective committed effective dose.

7. Stochastic Effect
•The probability of occurrence increases with increasing absorbed dose but the
severity in affected individuals does not depend on the magnitude of the
absorbed dose.
• All-or-none phenomenon,
•e.g cancer or genetic effect.
•Probability of such effects occurring increases with dose, their severity does not.
•There is no threshold, that is, no dose below which the probability of an effect is
zero.
• The dose-response relationship is therefore linear, or linear-quadratic, with no
threshold.

8. Deterministic effects
•Increases in severity with increasing absorbed dose in affected individuals, owing to
damage to increasing number of cells and tissues.”
• Examples : organ atrophy, fibrosis, lens opacification, blood changes, and decrease
in sperm count.
•The dose-response relationship is therefore sigmoid after a threshold

9. OBJECTIVE OF RADIATION
PROTECTION
(1) to prevent clinically significant radiation-induced deterministic effects
by adhering to dose limits that are below the apparent or practical threshold,
(2) to limit the risk of stochastic effects (cancer and hereditary effects) to a
reasonable level in relation to societal needs, values, and benefits gained.
The objectives of radiation protection can be achieved by
reducing all exposure to as low as reasonably achievable
(ALARA)

10. ALARA
•The recommendation that standard-setting committees would like to make for
personnel protection is zero exposure
•Taking into account:
Social,
Technical,
Economic,
Practical, and
Public policy considerations
•NCRP 116 ALARA Guidance:
Justification
The need to justify radiation dose on the basis of benefit
Optimization
the need to ensure that the benefits are maximized
Limitation
the need to apply dose limits

11. Six fundamental principle should be considered
 Eliminate or reduce the source of radiation,
 Contain the source,
 Minimize time in a radiation field,
 Maximize distance from a radioactive source,
 Use radiation shielding, and
 Optimize resources.
Hierarchy of Controls
•
Engineered Controls
•
Administrative controls
•
Personnel protective measures

15. NIRL
(Negligible Individual Risk Level ) :
•A level of average annual excess risk of fatal health effects
attributable to irradiation, below which further effort to
reduce radiation exposure to the individual is unwarranted
•NIRL should not be thought of as an acceptable risk level, a
level of significance, or a limit ,nor should it be the goal of
ALARA, although it does provide a lower limit for application
of the ALARA process.

16. Principle of radiation protection.
Dose received by individuals depends on
1. Time : Dose = Dose Rate X Time
2. Distance : The Inverse Square Law i.e Dose ∞ 1/distance2
3. Shielding:
Choice of shielding material and thickness depends on the
energy and intensity of the beam
Half Value Thickness (HVT):
Thickness of a specified material which reduces the intensity of radiation
to one half.
I = I0e -µt,
I0 (R/min) →Incident intensity
I (R/min)→transmitted intensity,
µ→attenuation coefficient

17. Tenth Value Thickness (TVT)
 The thickness of the shielding material which reduces the intensity to
one tenth of its original value is called TVT of the material for photon
beam
 TVT reduce intensity by a factor 1/10n
 1 TVT = 3.3 HVT
Thicknesses Of Some Materials, That Reduce Gamma Ray
Intensity By 50% (1/2) Include
Material
Halving
Thickness,
[inches]
Halving
Thickness,
[cm]
Density,
[g/cm³]
Halving Mass,
[g/cm²]
Lead
0.4
1.0
11.3
12
Steel
0.99
2.5
7.86
20
Concrete
2.4
6.1
3.33
20
Packed soil
3.6
9.1
1.99
18
Water
7.2
18
1.00
18
Lumber or
other wood
11
29
0.56
16
Air
6000
15 000
0.0012
18

18. Structural Shielding Design
•For protection calculations, the dose-equivalent limit is assumed to be 0.1
rem/week for the controlled areas and 0.01 rem/week for the noncontrolled
areas. These values approximately correspond to the annual limits of 5
rem/year and 0.5 rem/year, respectively.
•Protection is required against three types of radiation:
Primary Radiation,
Scattered Radiation
Leakage Radiation .
•Primary Barrier : sufficient to attenuate the useful beam to the required
degree.
• Secondary Barrier : against stray radiation (leakage and scatter).

19. BARRIER THICKNESS DEPENDS ON:
1 . Workload (W): terms of weekly dose delivered at 1 m from the source
2. Use Factor (U): Fraction of the operating time during which the radiation
under consideration is directed toward a particular barrier
3. Occupancy Factor (T): Fraction of the operating time during which the area
of interest is occupied by the individual
4 .Distance (d): Distance in meters from the radiation source to the area
to be protected. Inverse square law is assumed for both the primary and
stray radiation.

20. A. Primary Radiation Barrier:
P: maximum permissible dose equivalent for the area to be
protected(e.g., 0.1 rad/week for controlled and 0.01 rad/week for non
controlled area)
B : Transmission factor for the barrier to reduce the primary beam dose to
P in the area of interest
For megavoltage x- and γ radiation, equivalent thickness of various
materials can be calculated by comparing tenth value layers (TVLs) for
the given beam energy

21. B. Secondary Barrier for Scattered Radiation:
The amount of scattered radiation depends on:
1. the beam intensity incident on the scatterer
2. the quality of radiation
3. the area of the beam at the scatterer
4. the scattering angle
α : fractional scatter at 1 m from the scatterer, for beam area of 400 cm2
incident at the scatterer
Bs :
Transmission factor for scattered radiation
δ′ : The distance from the scatterer to the area of interest
F : area of the beam incident at the scatterer

22. C. Secondary Barrier for Leakage Radiation
•5 to 50 kVp: The leakage exposure rate4 shall not exceed 0.1 R in any 1 hour at any
point 5 cm from the source assembly.
•Greater than 50 kVp and less than 500 kVp.: The leakage exposure rate at a
distance of 1 m from the source shall not exceed 1 R in any 1 hour
•Greater than 500 kVp: The leakage dose rate from the source assembly at any point
at a distance of 1 m from the electron path between the source and the target shall not
exceed 0.5%
•Cobalt teletherapy:
Beam in the “off” position shall not exceed 2 mrad/h on the average and 10
mrad/h at a distance of 1 m from the source.
Beam in the “on” position, the leakage dose rate from the source housing shall
not exceed 0.1% at a distance of 1 m from the source.

23. D. Door Shielding
• Require a motor drive as well as a means of manual operation in case of
emergency
•With a proper maze design, the door is exposed mainly to the multiply
scattered radiation of significantly reduced intensity and energy
•In most cases, the required shielding turns out to be less than 6 mm of lead.

24. Ionising Radiations Regulations(IRR99)
Designation of areas
Controlled areas
Supervised Area
Uncontrolled Area
•For protection calculations, the dose-equivalent limit is assumed to be
0.1 rem/week for the controlled areas (5 rem/year )and
0.01 rem/week for the noncontrolled areas. (0.5 rem/year ).

25. Controlled areas
 Areas where a person is likely to receive an effective whole





body dose of more than 6mSv per year or where there is
significant risk of spreading contamination outside the
work area.
Must be physically demarcated
Must have suitable signage
Local rules should be drawn up
Radiation Protection Supervisor appointed
Environmental and personal monitoring should take place

26. Supervised areas
 Any area where the conditions need to be kept under review
 Any person is likely to receive an effective dose >1mSv/y or > than 1/10 of any
other dose limit.
 It does not automatically follow that outside every controlled area there will be
a supervised area.
Uncontrolled Areas
•All other areas in the hospital or clinic and the surrounding environment
• Uncontrolled areas are those occupied by individuals such as patients, visitors
to the facility, and employees who do not work routinely with or around
radiation sources.
• Areas adjacent to but not part of the x-ray facility is also uncontrolled areas.

27. Instruments for detecting and measuring
radiation
 Level of radioactive contamination
 Radiation dose rate in area
 Identity and quantity of radioactive
material
 Accumulated dose to individuals in
area
Survey meters
Laboratory counters
Personnel dosimeters
- 27

28. CONTD……
 Survey meters
•
•
•
Geiger-Mueller (GM) instruments
Ionization chamber instruments
Scintilation instruments
 Laboratory counters
 Personnel dosimeters
• Photographic film dosimeters (Film Badges)
• Thermoluminescent dosimeters
• Pocket dosimeters

29. Survey Meters
Survey meters are used to determine the extend of possible
contaminations.
Most frequently used is the
Geiger-Miller (GM) meter, which are
based on the ionization effects of
radiation in gas. The radiation is
completely absorbed in the counter
gas, creates a charged particles which
are collected in the field of the applied
voltage and converted to an electrical
pulse.
The number of pulses corresponds to the number of absorbed
particles, but is independent from the applied collection voltage.
Therefore the GM detector is used for measuring the rate of the
radiation not the absorbed dose (energy).

30. CD V-715 Civil Defense High-Range Survey Meter
0-500 R/hr range
3.25 pounds, die cast aluminum and drawn steel case, watertight, will
float. Powered with one D-sized battery, continuously for 150
hours, longer if on intermittent basis.
Instrument accuracy on any of its four ranges is within +- 20% of true
dose rate. Accuracy maintained throughout temperature ranges of 20 F to +125 F, relative humidities to 100% and altitudes up to 25,000'.

31. PERSONAL MONITORING DEVICES
1.Radiation film badges are composed of two pieces of film, covered by light
tight paper in a compact plastic container. Various filters in the badge holder
allow areas to be restricted to X-ray, -ray, -rays only.
For -radiation the sensitivity is in the range of 10 - 1800 mrem.
For -radiation the sensitivity is in the range of 50 - 1000 mrem.
2.Pocket dosimeter
The pocket dosimeter or pen dosimeter is a common small sized ion
chamber which measures the originated charge by direct collection on a
quartz fiber electroscope.

32. THERMOLUMINESCENT
DOSIMETER(TLD)
TLD is the primary form of personnel radiation monitoring dosimeter.
 Thermoluminescent dosimeters make use of the property of certain materials which
absorb energy when exposed to X , Gamma or Beta radiation
 On heating, the absorbed energy is released in the form of visible light. A plot of light
intensity emitted against temperature is known as a glow curve.
 For a given heating rate, the temperature at which the maximum light emission
occurs, is called the glow-peak temperature and it is characteristic, of the individual
TL material (also called phosphor)
 The quantity of the visible light emitted (TL output) is found to be proportional to the
energy absorbed by the TL material.
 The estimation of radiation exposure may be based either on the height of the glow
curve (differential method) or the area under the glow curve (integral method).

33. CONTD…………
•The TLD personnel monitoring system essentially consists of two major parts: TLD
badge and the TLD badge reader.
•The TLD Badge Reader comprises of a plastic cassette containing three Teflon TLD
discs (13.3mm and 0.8mm thick) that are mechanically clipped on to circular holes
(12.0mm) punched in an aluminium card (52 x 30 x 1mm).
•Materials :
calcium fluoride,
Lithium fluoride
calcium sulfate
lithium borate
calcium borate,
potassium bromide
Feldspar

34. •Three CaSO4: Dy Teflon TLD discs are mechanically clipped on an Alluminium plate.
An asymmetric “V” cut is provided in the card to ensure its loading in the plastic
cassette
TLD CARD
Dimensions of
A1 card: 52.5 mm X 30.0 mm X 1.0 mm
Hole on A1 Plate : 12.0 mm dia
Dimensions of TLD Disc : 13.3 mm dia
Three well-defined regions in the plastic cassette / holder corresponding to three TLD
discs of the TLD card.
i. Disc D1- sandwiched between a pair of filter combination of 1.0mm thick Cu
(Copper filter nearer to the
disc).
ii. Disc D2- sandwitched between a pair of 1.6mm thick (180mg/cm2) plastic filters
and
iii. Disc D3- under a circular open window.

35. TLD CASSETTE DIMENSIONS
In this design of the TLD cassette, dimension of some of the filters was altered and
crocodile clip was replaced by a smaller size clip. The cassette was made of ABS plastic
(white) and filters were embedded into the plastic body.
Dimensions :
Main Body :
Cu filter (rectangular) : 32 mm x 16 mm x 1mm
A1 filter (circular) : 13.6 mm x thickness – 0.6 mm
Plastic filter (rectangular) : 30.5 mm x 21 mm x 1.6 mm
Open window : Dia – 14.5 mm
Slider part :
Cu filter (rectangular) : Dia – 15.6 mm, x thickness – 1 mm
A1 filter (circular) : Dia – 12.6 mm, x thickness – 0.6 mm
Plastic filter (rectangular) : Dia – 25 mm, x thickness – 1.5 mm
Open window : Dia – 13.5 mm

36.  There are two types of TLD badges/ cassettes in use namely,
1. Chest Badge for whole body monitoring and
2. Wrist Badge
 For extremity dosimetryAccuracy of measurement is about ± 20%.
 TLD provides to us by BARC measures doses due to X-rays, gamma rays
and beta rays.
 It can be reused.
 Disadvantage → it is not a permanent record, once heated for
measurement, it lost all its data that recorded during radiation exposure.

37. Personal radiation monitoring device(prmd)
http://www.epa.sa.gov.au/xstd_files/Radiation/Guideline/guide_prmd.pdf

38. Personal Protective Equipment (PPE) in
Radiation Emergencies
•In a radiation emergency, the choice of appropriate personal protective equipment
(PPE) depends on
•Response role and specific tasks
•Risk of contamination
•PPE can protect against
•External contamination
•Internal contamination via inhalation, ingestion, absorption through open
wounds
•Other physical hazards (e.g., debris, fire/heat, or chemicals)
•PPE should include a personal radiation dosimeter whenever there is concern about
exposure to penetrating ionizing radiation.
•Direct-reading dosimeters should be worn so that a worker can easily see the
read-out and/or hear warning alarms.
•Recommended respiratory PPE includes a full-face piece air purifying respirator with a
P-100 or High Efficiency Particulate Air (HEPA) filter3.

39. RADIATION SAFETY OFFICER (RSO)
Responsibility :
1. Recommending or approving corrective actions,
2. Identifying radiation safety problems,
3. Initiating action, and ensuring compliance with regulation
4. Assisting the Radiation Safety Committee
Duties :
•Annual review of the radiation safety program for adherence to ALARA concepts.
•Quarterly review of occupational exposures.
•Quarterly review of records of radiation level surveys.
•Educational Responsibility
•Briefings and educational sessions to inform workers of ALARA programs.
•The RSO will ensure that authorized users, workers, and ancillary personnel who
may be exposed to radiation will be instructed in the ALARA philosophy
•Establishment of investigational levels in order to monitor individual occupational
external radiation exposures

40. Regulations that govern the use of cobalt-60 as teletherapy
1. Maintenance and repair
2. License amendments
3. Safety instructions
4. Safety precautions
5. Dosimetry equipment
6. Full calibration
7. Periodic spot checks
8. Radiation surveys
9. Five-year inspection

41. Protection for the Cobalt room
 Treatment room must be equipped with a radiation monitor which
should be clearly visible from the door
 CCTV : Monitor the patient movement during treatment.
 Interlock should be provided at the door
 Emergency switches must be provided

42. In Case of radiation emergency:
SOURCE DRAWER T-BAR
 When yellow colored portion of the T – bar entirely inside the head
cover, the source is in the fully shielded position.
 If the amber colored portion of the T-bar is visible and the red colored
portion is entirely inside the cover, radiation fields can be considered as
safe.

43. RESPONSIBILITIES OF RADIATION SAFETY OFFICER IN
CASE OF ACUTE EMERGENCY
 To check radiation level in the maze area with
Survey meter.
 If the red tip of the source drawer position
indicator rod is visible, high radiation fields
will be present.
 To take T-bar, from its location, enter room
but avoid exposures to the treatment beam.
 Insert the end of the T-bar over the red
indicator rod through the head cover.
 Apply firm pressure to the T-bar and push the
source back into the fully shielded position.
Insert the locking pin.

44. Radiation safety in HDR Remote after loading
unit
 Source safe (container) in HDR unit
 Source movement and drive system
 Manual retraction of sources
 Emergency safe container
 Source transfer container

45. In Treatment Room (HDR)
 The radiation monitor( Zone monitor)
 CCTV to monitor patient during treatment
 Intercom system to communicate with the patient
 Electrical interlocks at the entrance.
 Portable shield / area monitor
 Emergency stop motor / battery back up
 Manual retraction hand crank
 Posted emergency procedures

46. Summary of Atomic Energy Act 1962
 Atomic Energy Act promulgated in 1962 by Central Government.
 Power to control over radioactive substances or radiation generating plant
in order to prevent radiation hazards and safety of radiation workers and
public.
 Power to inspect to ensure compliance of the act and rules made their
under.
 Power to appoint competent authority
 Power to make rules.
Under section 27 of the Atomic Energy Act. 1962, Atomic
Energy Regulatory Board (AERB) was constituted on
November 15, 1983 to carry out certain regulatory and safety
functions envisaged under section 16, 17 and 23 of the Atomic
Energy Act 1962.

47. Functions of AERB
 Develop safety codes, guides, and standards
 Review operational experience in the light of the radiological and other safety
criteria recommended by International bodies and evolve major safety
policies.
 Prescribe acceptable limits of radiation exposure.
 Review the emergency preparedness plans
•
•
•
Promote research and development.
Prescribe the syllabi for training of personnel.
Enforce rules and regulations promulgated under the Atomic
Energy Act, 1962 for radiation safety in the country.
•
Maintain legislation with statutory bodies in the country as well as
abroad regarding safety matters.
•
To keep the public informed on major issues of radiological safety
significance.

48. THANK YOU