Radiation Protection:
Principles, Methods and
Monitoring
Moderator: Mr Anoop Bharadwaj

Retrospective
• “..the surgeons of Vienna and Berlin believe that the
Roentgen photograph is destined to render inestimable
services to surgery.”
– JAMA, 1896
• “(Radiation is) the most serious agent of pollution of the
environment and the greatest threat to man’s survival on
earth.”
– E.F. Shumacher, in Small is Beautiful, 1973

Introduction
• The acceptance by society of the risks associated with
radiation is conditional on the benefits to be gained from the
use of radiation.
• Nonetheless, the risks must be restricted and protected
against by the application of radiation safety standards.
• It is therefore essential that activities involving radiation
exposure be subject to certain standards of safety in order to
protect the individuals who are exposed to radiation:
– Occupationally,
– For medical diagnostic or therapeutic purposes
– As members of the public.

Radiation Effects

Effects of Radiation Exposure
Radiation Exposure
Stochastic Effects / Non stochastic effects/
Probablistic Deterministic
➢ Have no threshold levels of ➢ Have definite threshold levels
radiation dose. of radiation dose.
➢ The probablity of the effects ➢ The probablity of the effects
is proportional to the dose. is proportional to the dose.
➢ A latent period is seen ➢ A latent period is seen
between the time of between the time of
exposure and the events to exposure and the events to
manifest manifest
➢ Severity independant of dose ➢ Severity may be proportional
received to the dose received.
➢ Seen when the cells are ➢ Seen when the cells are killed
modified rather than killed. or loose capability to divide.

Effects if Radiation Exposure
Deterministic
Stochastic, Quadratic
Stochastic, Linear

Stochastic Effects
• These are primarily of two types:
– Carcinogenesis
– Hereditary effects
• Both have:
– A random nature of appearance.
– No threshold dose for appearance.
– Definite latent period for appearance after exposure.
– Probablity of induction increases with the dose received.
– Severity of the effect is independent of the dose received.
– A risk that can be defined on epidemiological studies only.

Radiation Induced Carcinogenesis
• Data from BIER V (US population) and the UNSCEAR
(Japanese Atomic Bomb survivors) show that:
– Relative risk of radiation induced cancers is a linear function of
doses upto 2 Sv
– At lower dose range of 0 -0.5 Sv risks are slightly higher than
the extrapolated risk.
– Risk of radiation induced cancers varies with age with patients
at younger age being more susceptible.
– Females < 15 years are most susceptible.

Risk Estimates for Carcinogenesis
• Cancer accounts for ~ 25% of all deaths in developed
nations
• The annual risk of fatality from cancer is shown in the table
below.
• ICRP estimates that for each radiation induced cancer 13 -15
years of life will be lost (but most will occur at ages of 68 –
70 yrs)
Population High Dose, High Dose Low Dose, Low Dose
Rate Rate
Working Population 8 x 10 per Sv 4 x 10 -2 per Sv
-2
Whole Population 10 x 10 -2 per Sv 5 x 10 -2 per Sv

Radiation induced Hereditary effects
• Radiation induced hereditary effects are secondary to
mutations which are passed on to the progeny
• Radiation doesnot cause new types of mutations but
increases the frequency of naturally occuring mutations
• Three classes of hereditary effects are known:
– Gene mutations
– Chromosomal abberations ( Down's Syndrome)
– Multifactorial (Neural tube defects)
• ICRP estimates that risk of hereditary effects due to radiation
exposure is:
– 0.2% per Sv in general population
– 0.1% per Sv in working population

Radiation induced Mutations
• Radiosensitivity of different mutations vary widely – so an
average mutation risk is considered.
• Low dose rate radiation is less effective in inducing
mutations
• Time interval between exposure and conception plays an
important protective role.
– A period of 6 months is therefore recommended between
radiation exposure and conception.
• Radiation induced mutations can be transmitted across the
generations.
• Average “doubling dose” for humans is considered to be 1.56
Sv

Risk estimates of Hereditary effects
Disease Class Base Frequency First genera- 2nd generation
per million live tion risk per risk per million
births million live live births
births
Mendelian mutations 16500 750 – 1500 1300 – 2500
Chronic multifactorial 65000 250 – 1200 250 -1200
diseases
Congenital abnormal- 60000 2000 2400 – 3000
ities
Total 738000 3000 – 4700 3950 – 6700
Total risk per Gy NA 0.41 – 0.64 % 0.53 – 0.91 %

Fetal Effects
• Radiation risks to fetus are related to:
– Exposure magnitude
– Time of pregnancy
• Radiation risks are most significant during
organogenesis and in the early fetal period.
• Threshold for malformations:
– 100 – 200 mGy (Malformations)
– 100 mGy (Mental Retardation): Risk coeffecient
is 0.4 per Sv
• A exposure 1mSv is safe for a fetus – normal
exposure from background radiation
• A dose > 0.1 Sv is considered the threshold
beyond which an MTP should be considered.

Fetal Effects
• Time of radiation vs effect:
– 2 – 3 weeks: Most embryos are aborted
– 4 – 11 weeks: Severe abnormalities in most organs
– 11 – 16 weeks: Mental retardation and stunting more common
– 16 – 25 weeks: Mild degree of mental retardation and
microcephaly
– > 30 weeks: Usually leads to functional disabilities in later life
• Carcinogenesis:
– Most exposures implicated occur in 3rd trimester
– Doses more than 1 mSv will increase the risk – but no
threshold apparent.
– Excess absolute risk – 6% per Sv.

ICRP quantities for Stochastic effects
• The concept of detriment as recommended by the ICRP for
stochastic effects includes the following quantities:
– The probability of fatal cancer attributable to radiation
exposure;
– The weighted probability of incurring a non-fatal cancer
– The weighted probability of severe hereditary effects
– The length of lifetime lost, if the harm occurs.

Safety Standards
• These are defined based on the knowledge of the knowledge
of effects of radiation and the principles of radiation
protection.
• The UNSCEAR (United Nations Scientific Committee on the
Effects of Atomic Radiation, 1955) is the central body which
collects the data regarding this.
• The IAEA uses the data from ICRP in designing the safety
standards.
• Consensus guidelines on these safety standards are
published as International Basic Safety Standards for
Protection against Ionizing Radiation and for the Safety of
Radiation Sources (referred to as BSS)
• Latest version – 1996 (IAEA safety series 115)

Types of Radiation
Exposures

ICRP Reference Man
• The concept of a ‘reference human’ to help manage the
many different situations in which human beings would or
could be exposed to ionising radiations.
• ICRP published comprehensive report on Reference Man in
1975 (ICRP, 1975) – recent revision is in ICRP 2002.
• The purpose of Reference Man was:
– Create points of reference for the procedure of dose
estimations to humans.
– Derivation of relevant quantities and units for their
interpretation
– Considering the relationships between doses to different parts
of the human body and their effects.
• Most countries have modified the concept of reference man –
e.g. Indian Reference Man

ICRP Reference Man
• ICRP 89 has given definitions and data for reference humans
of both sexes in 6 different age groups:
– Newborn
– 1 Yr
– 5 Yr
– 10 Yr
– 15 Yr
– Adult
• Reference man is defined as being between 20-30 years of
age, weighing 70 kg, is 170 cm in height, and lives in a
climate with an average temperature of from 10 to 20 C. He
isa Caucasian and is a Western European or North American
in habitat and custom.”
• Recently weight has been revised to 73 Kgs.

Indian Reference Man
*Jain S.C., Mehta S.C., Kumar B., Reddy A.R. and Nagaratnam A . (1995) Formulation of the Reference Indian
Adult: Anatomical and Physiological Data. Health Phys. 68, 509-522.

Types of exposure
• 2 types of exposure have been defined by the ICRP:
– Normal exposures
– Potential exposures
• Humans can be exposed to radiation while:
– Persuing their normal occupation (Occupational exposure)
– Undergoing some medical or dental procedures for diagnosis
or evaluation (Medical exposure)
– During normal daily life (Public exposure)

Normal Exposures
• These are defined as the radiation exposure that will occur as
a matter of course in the industry or in the medical
procedures.
• The exposure will have a predictable magnitude albiet with
certain degree of uncertainity.
• Controlling the doses delivered in these normal exposures
is the method specified by IAEA.

Potential Exposures
• These are exposures that are feasible but not certain to
occur.
• These can become actual exposures if there is equipment
malfunction, design flaws of operating failures.
• The primary means for controlling potential exposures is by
optimizing the design of installations, equipment and
operating procedures with the following aims:
– To restrict the probability of occurrence of events that could
lead to unplanned exposures
– To restrict the magnitudes of the exposures that could result if
such events were to occur.

Occupational Exposures
• Occupational exposure which is defined as all exposures
of workers incurred in the course of their work (with the
exception of exposures excluded from the BSS and
exposures from practices or sources exempted by the BSS).

Medical Exposures
• Medical exposure which is defined as exposure incurred:
– By patients as part of their own medical or dental diagnosis
or treatment
– By persons, other than those occupationally exposed,
knowingly while voluntarily helping in the support and
comfort of patients
– By volunteers in a programme of biomedical research
involving their exposure.
• Three broad classes:
– Patients
– Caregivers
– Medical test subjects

Public Exposure
• Public exposure, which is defined as exposure incurred by
members of the public from radiation sources,
– Excludes any occupational or medical exposure and the
normal local natural background radiation
– But includes exposure to authorized sources and practices
and from intervention situations.

Radiation Protection
Quantities

Quantities for radiation protection
• The absorbed dose is the basic physical dosimetry quantity,
but it is not entirely satisfactory for radiation protection
purposes because the effectiveness in damaging human
tissue differs for different types of ionizing radiation.
• So other qunatities that have been proposed to quantify the
biological and physical effect of the radiation dose are:
– Organ dose (Mean Physical Dose, DT)
– Equivalent dose (Modifier for radiation type, HT)
– Effective dose (Modifier for organ type, E)
– Committed dose (Modifier for the time of exposure)
– Collective dose (Modifier for the population exposed)

Organ Dose
• The organ dose (DT) is defined as the mean dose in a
specified tissue or organ of the human body.
• Unit is cGy / Kg or joules / Kg

Equivalent Dose
• The extent of biological effect produced by radiation depends
on:
– The physical organ dose
– The characteristics of the radiation beam employed.
– The pattern of dose deposition.
• The equivalent dose (HT) is the organ dose multiplied by a
radiation weighting factor wR to account for the effectiveness
of the given radiation in inducing health effects.
If more than one type of radiation is
used.

Equivalent Dose
• The SI unit for equivalent dose is joules/kg or sievert
• One Sievert = 100 rem (radiation equivalent man)
• Older ICRP recommendations used the quantity of dose
equivalent (H)
• This was the dose to the a point in the organ multiplied by
the radiation weighting factor (wr).
• Radiation Weighting factor is a somewhat arbitrarily
chosen conservative value based on a range of RBEs related
to the linear energy transfer (LET) of the radiation.

Radiation Weighting Factors
Radiation Type Weighting Factor
Photons all energies 1
Electrons all energies 1
Protons, other than fission fragments, energy more than 2 MeV 5
Neutrons energy < 10 KeV 5
10 KeV to 100 KeV 10
100 KeV to 2 MeV 20
> 2 MeV to 20 MeV 10
> 20 MeV 5
Alpha particles, fission fragments, heavy nuclei 20
1) Quality factors for various radiation types; Source: ICRP publication 60
2) Excludes Auger electrons
3) All values relate to radiation incident on the body for external radiation and for
internal sources, emitted from the source

Effective Dose
• The probablity of stochastic effect in a tissue depends on the
type of tissue irradiated.
• For the same equivalent dose the probablity of stochastic
effects in two organs will vary – so the concept of effective
dose.
• Previously the concept of Effective dose equivalent was used
which defined the effective dose to a point (as in Dose
Equivalent)

Effective Dose: Definition
• The Effective dose (E) is defined as the summation of
tissue equivalent doses, each multiplied by the appropriate
tissue weighting factor (wT), to indicate the combination
of different doses to several different tissues in a way that
correlates well with all stochastic effects combined (ICRP
Publication 60)
• The unit for effective dose is also Joules / Kg or Seivert (Sv)

Effective Dose: Tissue Weighting Factors
• Tissue weighting factors wT should represent the relative
contribution of an organ or tissue to the total detriment
due to the effects resulting from a uniform irradiation of
the whole body.
• They represent the likely proportionate risks of stochastic
events when tissues are irradiated.
• For low doses, individual organ or tissue detriments can be
treated as additive and the total detriment to the whole
body is the summation of individual detriments.
• For the whole body wT = 1 ; so a uniform equivalent whole
body radiation dose has a effective dose E = 1 x HT = HT

Tissue Weighting Factors
Tissue Weighting Factors For purposes of calculation, the
remainder is composed of the following
Gonads 0.20
additional tissues and organs:
Breast 0.12
Adrenals
•
Red Bone Marrow 0.12
Brain,
•
Colon 0.12 Upper large intestine,
•
Lung 0.12 Small intestine,
•
Kidney,
Stomach 0.05 •
Muscle,
•
Breast 0.05
Pancreas,
•
Bladder 0.05 Spleen,
•
Liver 0.05 Thymus and
•
Esophagus 0.05 Uterus
•
These organs are choosen as they are
Thyroid 0.05
likely to be selectively irradiated.
Skin 0.01
Bone Surface 0.01
Remainder 0.05
Total 1.00
Source ICRP Publication 60

Tissue Weighting factors
Those exceptional cases in which a single one of the remainder
tissues or organs receives an equivalent dose in excess of the
highest dose in any of the twelve organs for which a weighting
factor is specified, a weighting factor of 0.025 should be applied
to that tissue or organ and a weighting factor of 0.025 to the
average dose in the rest of the remainder as defined above.

Effective dose: Use and Limitations
• Effective dose is a useful concept as:
– It measures the degree of harm from a given dose of
radiation
– It can be used to compared different types of radiation
– Can be used to compare the dose from various types of
exposures.
• Limitation:
– Tissue weighting factors are defined based on flawed
estimates of stochastic effects
• For the whole body the ICRP recommends the use of
Effective doses for defining safety standards e.g. annual
dose limits.
– For individual organs and extremities, however equivalent
doses are used.

Committed Dose
• Useful concept for nuclear medicine
• Relates to the time which a radionuclide will remain inside
the body.
• It is defined as the specific time integral of the dose received
by the body after exposure to a specific radionuclide.
• A committed effective dose and equivalent dose may also be
defined by multiplying the tissue and radiation weighting
factors with the committed dose.

Collective Dose
• The collective dose relates to exposed groups or populations
• Is defined as the summation of the products of the mean
dose in the various groups of exposed people and the
number of individuals in each group.
• Can be defined for effective (E) or equivalent doses (H).
• The unit of collective dose is the man-sievert (man-Sv).
+ +
E1 x 6 E2 x 4 E3 x 9

Operational Quantities
• The organ dose (DT), equivalent dose (H) and effective dose
(E) are not directly measurable and there are no
laboratory standards to obtain traceable calibrations for the
radiation monitors using these quantities.
• For this reason, the ICRU has defined three measurable
operational quantities for protection purposes:
– Ambient Dose equivalent
– Directional Dose equivalent
– Personal Dose equivalent

Ambient Dose Equivalent
• The ambient dose equivalent at a point in a radiation field is
the dose equivalent that would be produced by the
corresponding aligned and expaned radiation field in the
ICRU sphere at the depth d at the radius opposite to the
direction of the radiation field.
– The ICRU sphere is a 30 cm diameter tissue equivalent
sphere composed of:
• 76.2% Oxygen
• 11.1% Carbon
• 10.1% Hydrogen
• 2.6% Nitrogen
– depth d = 10 mm is recommended for penetrating Radiation.

Directional Dose Equivalent
• The directional dose equivalent at a point in a radiation field
H (d,Ω) is defined as the dose equivalent that would be
produced by the corresponding expanded field in the ICRU
sphere at depth d on a radius in a specified direction Ω.
• For weakly penetrating radiation the depth of measurement
recommended is 0.07 mm.

Ambient and Directional Dose equivalent
a Radiation
Field
d
ICRU Sphere

Personal Dose Equivalent
• The personal dose equivalent Hp(d) is defined for both
strongly and weakly penetrating radiations as the equivalent
dose in soft tissue below a specified point on the body at an
appropriate depth d.
• The relevant depth is:
– 10 mm for photon energies above 15 KeV
– 0.07 mm for photon energies below 15 KeV and β particles in
skin
– 3 mm for photon energies below 15 KeV and β particles in the
lens
• The personal dose equivalent from exposure to
penetrating radiation during the year is the radiation quantity
to be compared with the annual dose limits (for effective
dose).

Background Radiation
Exposure

Background Radiation
• The total effective dose equivalent for a member of the
population in the United States from various sources of
natural background radiation is approximately 3.0 mSv/year
(300 mrem/year).
• The sources of background radiation include:
– Cosmic Radiation
– Terrestrial Radiation
– Internal Radioactive nuclei
• The estimated total annual exposure is estimated to be 3
mSv.

Natural Radiation: Estimated Doses
Lung Gonads Bone Surfaces Bone Marrow Other Tissue Total
W 0.12 0.25 0.03 0.12 0.48 1
T
Cosmic 0.03 0.07 0.01 0.03 0.13 0.27
Cosmogenic 0 0 NA 0 0 0.01
Terrestrial 0.03 0.07 0.01 0.03 0.14 0.28
Inhaled 2 NA NA NA NA 2
In the body 0.04 0.09 0.03 0.06 0.17 0.4
Estimated total effective dose equivalent rate for a member of the population in the United States and
Canada from various sources of natural background radiation

Radiation Protection
Principles

Basic Principles of Radiation Protection
Radiation
Protection
Dose
Justifiable ALARA
Limits
Exposure

Protetion Principles: Justifiable Use
A practice that entails exposure to radiation should only be
adopted if it yields sufficient benefit to the exposed
individuals or to society to outweigh the radiation detriment
it causes or could cause (i.e. the practice must be justified).
• Example:
– Benefit gained from the use of radiation therapy for treatment
of Ptyrisasis versicolor or thyroid goitre doesnot allow the use
of therapeutic radiation for these treatments
– Use of x-ray pelvimetry has significant hazaards for the fetus
so is no longer used routinely although may be useful for the
rare mother with some pelvic deformity

Protection Principles: Dose Limits
• Individual doses due to the combination of exposures
from all relevant practices should not exceed specified dose
limits for occupational and public exposure.
• Different dose limits are specified for the radiation workers
as the expected benefit from the work they do while handling
radiation will outweigh the small increase in risk.
• Pregnant radiation workers have to be protected so that the
fetus or embryo is given the same radiation protection as
given to that of the public.
– Dose limits are not relevant for control of potential exposures
– nor are they relevant for descisions on whether or how to
undertake an intervention.
– Dose limits are not applicable to medical exposure as the
benefit gained from the exposure outweighs the harm.

Protection Principles: Dose Limits
• Examples:
– Individual dose for a person who is delivering a letter to the
Co60 room would be counted in the quantum of radiation
exposure he had but not the radiation exposure after an x-ray
he had to have for the ankle after he slipped on the wet floor
while going out.
– If a pregnant female is given radiation for a therapeutic
indication then the dose to her is not counted but dose to the
fetus is counted.
– When a radiation worker becomes pregnant care should be
taken that during the period of gestation the exposure to
the fetus doesnot exceed that of a general person – note the
limits no longer apply after gestation unless she is working in
nuclear medicine where radionuclides can be absorbed by the
body – hence can be applicable for lactation.

Protection Principles: ALARA
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 be As Low As Reasonably Achievable (ALARA),
economic and social factors being taken into account,
and the doses they deliver and the risk they entail be
constrained (i.e. protection and safety should be
optimized.)

Protection Principles: ALARA
In diagnostic medical exposure, optimization of protection is
•
achieved by keeping the exposure of patients to the minimum
necessary to achieve the required diagnostic objective
– Thus in Obsteterics Xrays or CT Scans is better replaced with
clinical examination, USG or MRI.
– A skeletal survey is thus used for multiple myeloma but not
for other conditions like TB.
In therapeutic medical exposure, optimization is achieved by
•
keeping exposure of normal tissue ALARA consistent with
delivering the required dose to the planning target volume (PTV).
As a thumb rule a value of $ 1000 / 10mSv reduction in
•
exposure is considered reasonable. ( $ 10,000 in some special
situations too.)

Radiation Protection
History and Evolution

Radiation Protection: Dawn
• 1904: William H. Rollins, a Harvard-trained physician and a
practicing dentist, reported that the hazards of x-rays could
be reduced by using simple shielding methods.
• 1913: German Roentgen society developed 1st guidelines to
shield x-ray technicians from excessive exposure.
• 1928: The 2nd International Congress for radiology
established the International X-ray and Radium protection
committee.
– 1929: American Medical Association condemned the use of
radiation to remove hair.
– 1932: American Medical Association withdrew radium from
the list of remedies used for internal administration.

Radiation Protection: Tolerance Doses
• 1934: The International X-ray and Radium Protection
Committee along with it's American counterpart Advisory
Committee on X-ray and Radium Protection gave a
qunatitative tolerance dose of external radiation.
– Based on research on time to cause erythema
– Recommended limits:
• 0.1 R per day for the whole body
American Committee
• 5 R per day for the fingers
• 0.2 R per day whole body limit – International Committee
– The discrepancy was the result of the difference in rounding
off of similiar data by the two committes.
• 1941: The American committee defined the tolerance dose
from internal deposition of radium and radon.

The Birth of the Atomic Age

Radiation Protection in the Atomic Age
• Became more important as:
– Increasing use in the public, health and military domains
made radiation exposure more common
– Different aritficial radio-isotopes were generated with new
characteristics.
– Protection no longer was confined to radium and xrays.
• The American Committee was changed to National
Committee on Radiation Protection (NCRP) in 1946.
• The international Committee was changed to International
Committee on Radiation Protection (ICRP) similiarly.
• Based on the work of the geneticist H.J.Muller who had
indicated that the reproductive cells were vulnerable to even
smallest doses of radiation the concept of tolerance doses
were abandoned.

Permissible Dose
• The concept of tolerance dose indicated that there was a
level of radiation below which it was safe.
• The concept of stochastic effects of radiation invalidated this
dogma
• Most scientists rejected that there was a threshold below
which exposure to radiation was biologically innoncous.
• The concept of permissible dose was therefore introduced.

Maximum Permissible Dose (MPD)
• The NCRP defined the Maximum Permissible Dose as that
dose which, in the light of present knowledge is not
expected to cause appreciable bodily injury to the person
at any point during his lifetime.
• 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 – probablity 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.

Maximum Permissible Dose: NCRP
• The NCRP initially defined a dose level of 0.3 R per six day
work week for whole body external radiation exposure
(agreed on in 1948; published in 1954).
• Also proposed a detailed set of recommendations for
permissible levels of internal exposure from radio-isotopes
taken in the body.
• To provide an adequate margin of safety, it proposed
permissible levels as low as 1/10th of the numerical values
derived from the sketchy data then available.
• A seperate limit of 1/10th the adult dose (0.03 R/week) was
proposed for the minors < 18 yrs age.

Maximum Permissible Dose: ICRP
• The ICRP also used the proposal of MPD as given by the
NCRP and used the same dose limits (0.3 R/week).
• The one major area of difference was that it proposed the
dose of 1/10th the level of exposure for the radiation worker,
for the general public (0.03 R/week).
• This level was first given in 1953, to account for the risk
that the general population had from accidental exposure to
radiation.
• The ICRP's effort was the first attempt to produce radiation
protection guidelines for the general population outside
controlled areas where radiation workers worked.

Fallout Controversy

Change in MPD levels
• Several changes were introduced by the NCRP and ICRP after
the 1960s to account for the increased exposure from the
radioactive fallout from above ground nuclear bomb
explosions.
– Reduced the permissible doses for whole body exposures by
1/3rd.
– Introduced the rad as the unit for dose rather than roentogen
– Also introduced the term rem – radiation equivalent man – for
chronic low level radiation exposure
– The concept of RBE – Relative Biological Effectiveness – was
introduced to account for the difference in the biological
effects produced by different types of radiation.

Maximum Permissible Doses: ICRP
• In 1956, the ICRP published the revised limits for MPD in
terms of rem
– Maximum permissible dose limit for occupational exposure
was kept at 5 rem per year.
– Defined the MPD for the general population as 0.5 rem per
year.
– In addition defined dose levels that could be accumulated at
various ages:
• 50 rems to age 30 years
• 100 rems to age 40 years
• 200 rems to age 60 years.
– Thus the main thrust was providing additional protection for
the younger population who were considered to be at greater
risk for tranmiting genetic damage due to radiation.

Maximum Permissible Dose: NCRP
• The revised NCRP recommendations were published in 1958.
– Recommended average whole body radiation exposure from
external sources to 5 rem per year.
– Allowed a maximum limit of occupational exposure of 12 rem
in a year if
• Dose to person did not exceed the limit of 5 rems in the previous
year.
• Adequte past record of exposure existed.
– Also gave a formula to calculate the MPD as per the age:
• Accumulated MPD = 5 (Age in years - 18)
– Recommended that doses to persons living close to radiation
sources but outside controlled areas not to exceed 1/10th of
those proposed for occupational exposure.

MPD – Radionuclides
• In 1959 both ICRP and NCRP met and recommended the
permissible body burden of radionuclides.
• The recommended dose limits were:
– 0.5 rem per year for an individual (1/10th of the occupational
limit)
– 0.17 rem per year for the entire population ( 1/30th of the
occupational limit)

As Low as Practicable (ALAP)
• 1st proposed by Forrest Western, Director of the Division of
Radiation Protection Standards, AEC, USA (published in
1970)
• Applied primarily to low level radiation exposure due to
leakage from the Nuclear Reactors.
• Basically gave the axiom that the radiation exposure was to
be kept as low as possible practically from these reactors.
• The changes were introduced in an effort to tighten the limits
for effluent generated by the nuclear power plants – however
concrete dose limits could not be specified.

Radiation Protection
Modern Day

As Low As Reasonably Acheivable (ALARA)
• Initally defined by the NRC (1975) for the nuclear power
plants.
• Defined as:
– A level as low as reasonably achievable taking into account
the state of technology, and the economics of improvements
in relation to benefits to the public health and safety, and
other societal and socioeconomic considerations, and in
relation to the utilization of atomic energy in the public
interest.
• Primarily designed to make radiation protection for the
nuclear power plants economically acheivable
– Example: It is practical to shield a reactor inside a shield of
tungsten but not acheivable economically so concrete has to
be used.

Negligible Individual Risk (NIRL)
• Given by the NCRP only.
• Defined as the dose level at which the risk of annual fatal
health effects due radiation is so low that efforts to reduce
the exposure to the individual is unwarranted.
• The NCRP defined this limit as there was a need to define the
threshold of the lower risk below which efforts to reduce
the risk would not be warranted.
• However doesnot contravene the principles of permissible
dose limits and ALARA.
• The risk level of radiation induced fatal events choosen for
setting the NIRL is 10-7.
• This limit is specified for members of public only.

Rationale behind dose limits
• The risk for stochastic effects is taken into consideration
while designing the dose limits for occupational and public
exposure.
• These limits represent the dose levels at which the risk for
detriment are unacceptable (old reports had the dose levels
which were harmful or fatal).
• These represent the highest tolerable limits for continous
exposure.
• For the purpose of computations the risk coeffecients are
defined based on the risk per unit equivalent dose:
– Total risk coeffecient : 165 x 10-4 Sv-1
– Somatic risk component: 125 x 10-4 Sv-1
– Genetic risk component: 40 x 10-4 Sv-1 (Risk of severe genetic
damage over first 2 generations)

Modern Day Levels: ICRP
The maximum effective dose for occupational exposure is 50 mSv in 1 year and
the total exposure for 5 years should not exceed 100 mSv (avg 20 mSv).

Dose Limits: ICRP
The dose limits include:
•
– The effective dose from external sources in the specified period and
– The 50 year committed dose for internal sources (for children 70
year commited doses)
The 5 year limit is predefined – i.e. can't be defined after the
•
exposure has occured.
– Seperate limits are specified for the skin as it may be subject to
localized exposure – for skin, exposure averaged over an area of 1
cm2 is used irrespective of the area exposed.
– Eye/lens are defined as they make minimal contribution to the total
effective dose.
For the pregnant female the ICRP recommends a dose limit of 2
•
mSv to the women's abdomen.

Dose Limits: NCRP
Exposure Type Dose Levels
Occupational Exposure (Annual)
50 mSv
Whole Body Effective Dose Limits (Stochastic Effects)
Dose Equivalent Limits for organs (Non Stochastic Effects)
150 mSv
Lens of Eye
All others 500 mSv
Public Exposure (annual)
Effective dose euivalent limit, continous exposure 1 mSv
Effective dose euivalent limit, infrequent exposure 5 mSv
Dose Equivalent Limits for lens, skin and extremities 50 mSv
Embryo Fetus Exposures
5 mSv
Total dose equivalent limit
Dose Equivalent in a month 0.5 mSv
Annual effective education and training exposure 1 mSv
50 mSv
Dose Equivalent Limits for lens, skin and extremities
Source: The NCRP Report Number 91

Dose Limits : NCRP
• The NCRP has got certain guidelines that are not there in the
ICRP recommendations:
– Cumulative Dose: The numerical value of the cumulative
dose in terms of the tens of Sv exposure should not exceed
the numerical age of the person
– Example: For a 35 year old person the maximum cumulative
exposure is 350 mSv.
– For Students and apprentices below 18 yrs age, who are
exposed to radiation as a part of their education, the annual
effective dose limit is 1 mSv
– Guidelines for remedial action for public exposures:
• Effective dose limits > 5 mSv
• Exposure to radon or decay products > 0.007 J/m3

Difference between NCRP and ICRP
• Annual effective dose limits for occupational exposure:
– ICRP: 20mSv recommended, 50 mSv maximum allowed.
– NCRP: Allows 50 mSv annually.
• Risk defined for annual effective dose limits:
– ICRP: Both stochastic and nonstochastic events
– NCRP: For only stochastic events
• Cumulative Dose Limits for occupational exposure:
– In terms of contigous, predefined 5 year periods in ICRP.
– In NCRP the numerical calculation based on age is used.
• Organ Equivalent dose limits:
– Specified for lens, eye and skin in ICRP
– Specified for many organs in addition in NCRP

Difference between NCRP and ICRP
• Limits of Public exposure:
– 2 limits given by NCRP alone (ICRP limits correspond to the
limits of NCRP for continous or frequent exposure).
– ICRP max exposure limit in 1 year corresponds to NCRP
annual limits (5 mSv) for infrequent exposure.
• Fetal Dose Limits:
– In ICRP the total dose limit for fetus or embryo is 1 mSv (2
mSv to the lower abdomen of the pregnant lady).
– In NCRP the total dose limit for fetus or embryo is 5 mSv.
– In NCRP a monthly dose limit (1/10th the total limit) is also
specified.
• Threshold doses for remedial action and Negligible individual
annual risk levels:
– Defined by the NCRP only

Annual Limit on Intake (ALI)
• Radiation protection quantity defined for radionuclides which
can be ingested or inhaled.
• ALI is the smaller value of intake of a given radionuclide in a
year by the reference man that would result in a committed
effective dose equivalent of 5 rems (0.05 Sv) or a committed
dose equivalent of 50 rems (0.5 Sv) to any individual organ
or tissue.
– Thus in abscence of any external radiation exposure a
radiation worker can have 1 ALI of exposure from internal
isotopes
• Derived Air Concentration: The concentration of a given
radionuclide in air which, if breathed by the reference man
for a working year of 2,000 hours under conditions of light
work (inhalation rate 1.2 cubic meters of air per hour),
results in an intake of one ALI.

Comparision of risks
Occupation Mean Rate (per million per year)
Trade 40
Manufacture 60
Service 40
Government 90
Transport/ Public Utilities 240
Construction 320
Mines / Quarries 430
Radiation Worker* 20
Fatal Accident rates per million workers per year in
choosen occupations
* For radiation worker the 20 x 10-6 per year is the risk of any detriment not fatality

Organizations and Authorities
• The basic responsibility for radiation protection rests with the
authorized persons who will conduct the intervention causing
radiation exposure.
• The Goverment appoints a regulatory authority who will
enforce the rules of radiation protection
• In India, the regulatory authority is the AERB.
– Authorization granted for practice of radiotherapy is done via
liscences.
– The authorized person for the radiotherapy department is
therefore called a liscencee.
– Also provides services like personal dosimetry, equipment
calibration and maintainence of national standard dosimeter
(primary standard dosimeter).

Design of Radiation
Facilities: Principles

Principles
• 3 basic parameters influence the exposure that an individual
receives in a radiation field.
– Time: Longer the time spent in the radiation field greater the
exposure
– Distance: The exposure falls as a function of square of the
distance from the radiation source
– Shielding: Exposure can be reduced due to attenuation of the
priamry beam by shielding.
• Design of radiation facilities is basically design of the
shielding for a given set of time of exposure and distance.
• The design and amount of shielding will be a function of the
desired effective dose in a given area and the calculated
dose at that place in the event of no shielding.

Terminology: Types of Radiation
• Primary radiation: which is the radiation directly emitted
from the treatment machine through the collimator opening
in the case of external sources and from the radioactive
source in the case of brachytherapy.
• Scatter radiation: which is the radiation produced by the
scattering of the primary radiation beam from various media
struck by the primary beam, such as the patient, collimators,
beam shaping accessories and air.
• Leakage radiation: which is the radiation that escapes
through the shielded head of the therapy unit (for
accelerators leakage radiation only exists while the beam is
on; for cobalt units leakage radiation is always present).

Terminology: Barriers
• Primary Barrier: Where the primary radiation beam strikes
the wall it becomes a primary barrier. It includes the roof
and the floor if the radiation facility is situated above another
room.
• Secondary Barrier: These protect against the scattered and
leakage radiation.

Terminology: Areas
• Controlled areas: A controlled area is any area in which
specific protection measures and safety provisions could be
required for:
– Controlling normal exposures and preventing spread of
contamination during normal working hours
– Preventing or limiting the extent of potential exposures.
• Supervised area: An area for which occupational exposure
conditions are kept under review though protective
measures and safety provisions are not needed.
– Number and extent of controlled areas should be as limited as
possible as working in these areas require some form of
personal protection measures and administrative control.
– Controlled areas should be labelled appropiately and should
have restricted access.

Terminology: Workload and Use factors
• Workload: Refers to the typical radiation output from a
source per week at a well designated point.
• Use Factor: Refers to the number of times in a day the
radiation beam is directed towards the barrier in question
• Occupancy Factor: Refers to the amount of time the rooms
adjacent to the treatment room are occupied.
– The occupancy factor for an area should be considered as the
fraction of time spent by a single person who is there the
longest.
– The occupancy factor is best defined as the fraction of an 8 h
day or 2000 h year for which a single individual may occupy a
particular area.

Terminology: IDR and TADR
• When weekly dose limits are used instead of the
instantaneous dose the shielding requirements are often less.
– IDR refers to the direct reading of a dosimeter in dose per
hour averaged over one minute
– The IDR is an useful quantity for QA of the barrier after
construction.
– It is usually measured at a point 0.3 m from the barrier.
• Another quantity in this senario is the TADR – Time averaged
Dose Rate.
– TADR = IDR x ( Daily Beam on time / Length of working day)
– In UK an area of TADR2000 of ≤ 0.15 μSv/hr doesnot need
supervision
– US specifes a TADR limit of 20 μSv/hr in a public place.

General Design Guidelines
• Usually located at periphery of the hospital complex – avoids
the problem of therapy rooms being adjacent to high
occupancy areas.
• The ground level is preferred as the problem of shielding the
floor is less.
– If facility is located above the floor needs to be reinforced to bear
the weight of the machine in addition to the barrier.
• When ever possible the areas around a therapy machine
should be designated as a controlled area (including roofs
over the machines).
• Mazes should be designed whenever possible as they reduce
the need for a heavy shielded door.
• Doors should be provided at the maze entrance so that
casual entry of public can be avoided.

General Design Guidelines
• If a shieding barrier is required to reduce dose at the
entrance the door may be motorized. However in that event
manual operating facilities and several interlocks are needed.
• In ordinary dorros also interlocks are provided so that the
beam may be turned off when the door is opened.
• The control console should be provided with devices for
keeping a watch on the patient at all times.
– CCTV: Two cameras are recommended – 15 off and above the
gantry rotation axis for optimum patient viewing.
– Mirrors and door glass arrangement
– Lead glass for direct viewing – expensive, only for low energy.

General Design Guidelines
• Ducts for electrical materials should always run through a
secondary barrier and at an angle through it.
– No duct with diameter > 30 mm should penetrate the primary
barrier.
– The total length of the duct in the barrier should be greater
than the thickness of the radiation barrier.
– Ducts are best designed to follow the maze or reach the
treatment room through the ground.
• Warning signs, lights and audible alarms should be provided.
– Illumination signs may be two stage or three stage:
• 1st stage: When there is power in the treatment unit
• 2nd stage: When the treatment unit is programmed
• 3rd stage: When the beam is actually turned on.

Examples
Radiation on
indicator
Gamma Zone Monitor
Control Console

Examples
Steel Door with Trefoil
Lead Glass Window CCTV
Mirror

Brachytherapy Facilities
• Sources should be stored in a seperate lead lined chamber.
• Storage area for radium should be ventilated by a direct
filtered exhaust to the outdoors.
• The storage rooms are usually provided with a sink of
cleaning source applicators – a trap is provided to avoid
source loss.
• Source preperation bench should be provided with an L-
Bench with a lead glass window.
• Suitably long forceps should be used to provide as much
distance as practical between sources and the operator.
• Leak Test: A source is considered to be leaking if a
presence of 0.005 µCi or more of removable contamination is
measured

Design of Radiation
Facilities: Method

Building Materials

Building Material: Concrete
• Concrete density varies
according to the mixture.
• Concrete should be poured in
one go to prevent seams
from forming.
• Also during the mixing with
cement air may be trapped
so continous vibration of the
mix is required.
TVL of different materials used for shielding
• If preformed blocks are used,
heavy mortar is needed to
prevent shine paths.
Typical Barrier Thickness for different beam energies

Workload
• In EBRT, the typical number of patients treated in an eight
hour day is 50.
• NCRP Report 49 suggests a workload figure of 1000
Gy/week based on a dose of 4 Gy at 1 m per patient,
assuming a five day week for megavoltage facilities.
• For dual energy machines the same workload figure of 1000
Gy/week may be used. NCRP Report 51 suggests an
assumed workload of 500 Gy week for the higher energy,
with the remainder of the workload bein attributed to lower
energy X rays or electrons.
• If the workload is greater, the figures used for Gy/week
should reflect this.
So 50 petients per day , 5 days a week and 8 hours in a day is the basis for
workload calculations

Use Factors
• If conventional treatment techniques are to be used, NCRP
Report 49 suggests
– Use factors:
• 1 for the floor with the beam pointing vertically down.
• 0.25 for each wall and ceiling if specific values are not available.
– Use factors for brachytherapy, secondary barriers and doors is
considered as 1.
• These use factors may depend on the particular use of the
facility and also on the energy used.
• For example, a facility performing a large number of total
body irradiations may have a use factor greater than 0.25 for
one wall, and lower for other walls.

Occupancy Factors

Weekly Dose Rates
• ICRP 33 states that for design of shielding in radiation
facilities the actual dose values to individuals in the occupied
areas should be only 1/10th of the effective dose values
( 1/30th for the equivalent dose) used for calculation of the
shielding requirements.
– Using the dose limit of 0.01 mSv/ week at a distance of 4 m for 20
MV photons can increase the shielding requirement for the barrier
by 50 cm of RCC if designated as a public area as compared to what
can be used if it was designated as a controlled area.

Design Effective Limits
Design Limits USA UK
Occupational exposure Fraction of 10mSv in a year 6 mSv in a year
IDR is 7.5 μSv·h–1
Public exposure 1 mSv annually 0.3 mSv in a year
20μSv per hour IDR is <7.5 μSv·h–1
TADR is <0.5 μSv·h–1
TADR2000 <0.15 μSv·h–1

Weekly Dose Limits
• When caclulating the dose limits for shielding design a
number of conservative estimates are required:
– The attenuation of the beam by the patient is not considered.
– The maximum possible leakage radiation is assumed.
– The workload, as well as the use and occupancy factors, is
overestimated.
– An assumption is made that staff are always in the most
exposed place of the occupied area.
• For linacs producing X rays and electrons an assumption is made
that the linac always operates in the X ray mode.
• For dual energy linacs an assumption is made that the linac
always runs in the higher energy mode.

Installation Design

Thickness of Primary Barriers
• The thickness is back calculated using the desired dose
constraint.
• The formula used is:
P = Allowed dose per week outside the barrier
–
d = distance from the isocenter to the point outside barrier
–
SAD = Source to Axis distance (in m).
–
W = Workload (Gy / week at 1 m)
–
U = Use factor for that barrier
–
T = Occupancy factor or the fraction of time the area outside the barrier is likely to be
–
occupied.

Thickness of primary Barrier
• The value B directly gives the desired attenuation.
• This can be used to calulate the required thickness from the
next formula:
• An altenate method of calculation used is the use of IDR.

Secondary Barriers: Scattered Radiation
• The formula for calculation of the scattered radiation from
the patient is given as:
dsca is the distance from the radiation source to the patient, in m.
dsec is the distance from the patient to the point of interest, in m.
a is the scatter fraction defined at dsca.
F is the field area incident on the patient, in cm2.
The scatter primary ratio (a) is dependent on the energy of the X ray
beam and the scattering angle. These data are tabulated per 400 cm2
of irradiated field area for 60Co, 6, 10, 18 and 24 MV X ray beams.

Secondary Barrier: Scatterd Radiation
• The formula for calculation of the scattered radiation from
the walls is given as:
dw is the distance from the radiation source to the scattering
surface (wall), in m;
dr is the distance from the scattering surface (wall) to the point of
interest, in m;
α is the wall reflection coefficient, which depends on the wall
material, scattering angle, and beam energy (albedo factor)
A is the field area projected on the scattering surface (wall), in m2.

Secondary Barrier: Scattered Radiation
• For megavoltage beams, on the other hand, the maximum
energy of the 90-degree scattered photons is 500 keV.
• Therefore, the transmission of this scattered radiation
through a barrier is estimated to be approximately the same
as that for a 500-kVp useful beam.
• The use factor for the secondary barrier is considered unity.

Secondary barrier: Leakage Radiation
• The thickness of the secondary barrier to shield against
leakage radiation is given by the relationship
P is the design dose limit
ds is the distance from the isocentre to the point of interest in m
W is the workload
T is the occupancy factor.
The Use factor is always taken as 1.
As the use factor is one the distance of the source from the point of
concern is taken to lie at the isocenter as the average distance from all
gantry angles.

Secondary barrier: Leakage Radiation
• For a linear accelerator, national and international protocols
state that the leakage from the treatment head must not
exceed 0.5% of the primary beam, outside the useful beam
at 1 m from the path of the electrons between the gun and
target window and averaged over 100 cm2.
• In the plane of the patient, the leakage must not exceed an
average of 0.1% and a maximum of 0.2% over a 2 m radius
measured from the beam central axis.
• A simple quantity to keep in mind is the total dose from
leakage radiation should not exceed 1 mrad at 1 m per hour
from the gantry at any position.

Neutron Protection
• Considered for LINACs operating above 10 MV.
• Concrete has high water content – TVL for photoneutrons
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
• Slow neutron capture is enhanced by use of materials like
Boron.
– Slow neutron capture by Boron results in low energy gamma
ray production of 473 KeV.
– For doors 5% Boron is included in the polythelene in
treatment rooms.

Radiation Measurement
and Monitoring

Types of Monitoring
Radiation Monitoring
Personal Monitoring Workplace Monitoring
•For individual radiation •For the entire workplace and
worker the radiation rooms
•In form of radiation badges •Usually require some form of
•Worn on the person of the radiation detectors
worker •Integrated with the workplace
•Allows estimation of individual •Allows estimation of exposure
doses levels in the envioronment

Workplace Monitoring

Monitoring For Radiation Protection
• Types of instruments for Dose Measurements
– Dose rate meters used to measure the external exposure.
– Dosimeters which indicate the cumulative external exposure.
– Other types:
• Surface contamination meters which indicate the potential
internal exposure when a radioactive substance is distributed
over a surface.
• Airborne contamination meters and gas monitors which
indicate the internal exposure when a radioactive substance is
distributed within an atmosphere.

Dose Rate Meters
• A dose rate meter absorbs energy from penetrating
radiation.
• Capable of providing direct readings of the dose equivalent
rate in microsieverts per hour (mSv/h or mSv·h–1).
• Disadvantages
– These usually respond only to X, gamma and/or beta
radiations.
– Specialized instruments are necessary to measure neutron
dose equivalent rates.
– Dose rate meters may not be able to provide an accurate
response to rapidly changing or pulsed radiation fields.
Integrating dose rate meters and dosimeters are more
appropriate in such circumstances.

Dosimeter
• A dosimeter measures the cumulative energy absorbed as
a consequence of exposure to ionizing radiation.
– Personal dosimeters must be worn by radiation workers to
measure their radiation exposure.
– Passive dosimeters routinely monitor cumulative doses
resulting from an external exposure.
– Active dosimeters provide an immediate reading of the dose
in microsieverts (mSv) and may also provide an immediate
alarm signal when the measured dose approaches a value pre-
set by the manufacturer or user.

Principle of construction
• Detectors used:
Detector – Gas filled detectors
– Ionization chambers
– Proportional counters
– Geiger-Müller counters
Amplifier
– Scintillation counters
– Solid state detectors.
Processor
Display

Gas Filled Ionization Chambers
E1 and E2: Charged Electrodes
A: Housing for the gas
R : Radiation
P : Power Supply
D : Display

Practical Ionization Chambers
• Components:
– The cylindrical detector wall serves as the cathode (negative
electrode) and is normally made of air-equivalent, carbon
coated plastic or aluminium.
– The axial anode (positive electrode).
– 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 effeciency for high
energy photon radiation.
• Detector volumes of a few hundred cc are needed to
measure exposures in nanocoulombs per hour (nC·h–1).

Uses of Ionization chambers
• Simple Ionization Chambers: Can be used as personal
dosimeters and integrating dosimeters. They are most
commonly used for personal protection.
• Proportional Counters: Best used as surface
contamination monitors. Are sensitive to low energy photons
and beta particles.
• Geiger Muller Counters: These are commonly used for
dose and dose rate measurement.
• Scintillation Chambers: Primarily used for gamma,
electron and neutron spectrometry.
• Solid state detectors: Primarily used for gamma and
heavy particle spectrometry.

Neutron Detectors
• These consist of a ionization
chamber surrounded by a wall
made of polyethylene to
“thermalize” neutrons (M).
• The dosimeter is filled with a gas
like Boron trifluride (BF3) or He3.
• A perforated cadmium or
boronated plastic sheath (S) that
modifies the neutron energy so
that doses can be read in terms of
dose equivalent (directly
accounting for quality factor due
to different neutron energies).

Individual Monitoring

Individual Monitoring: Uses
• The exposures of the individual radiation worker needs to be
routinely monitored and records kept of their cumulative
radiation doses.
– They can also be used to retrospectively determine a dose
received by a worker.
– Individual monitoring is used to verify the effectiveness of
radiation control practices in the workplace.
– It is also used to detect changes in the workplace
– Confirm or supplement static workplace monitoring,
– Identify working practices that minimize doses and
– Provide information in the event of accidental exposure.

Principle of Personal Dosimetry
• Personal dosimeters are designed to measure the dose in
soft tissue at a defined depth below a specified point on the
body.
• The quantity personal dose equivalent Hp(d) is normally
determined at two depths:
– 0.07 mm : Accounts for the skin dose
– 10 mm*: Accounts for dose to blood forming organs.
• Personal dosimetry is required for:
– Workers who regularly work in controlled areas should have
their personal doses routinely assessed.
– Those who work full time in supervised area and/or
occasionally in controlled areas should also be considered as
potentia candidates for individual monitoring.

Film Badge Dosimeters
• Use a photographic film and several 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 fast film is responsive to high dose rates.
– The slow film is responsive to low dose rates.
• 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.

Film Badge Dosimeter: IAEA
W — window which allows all radiation which can
penetrate the wrapper to reach the film;
N — thin plastic filter which attenuates beta
radiation depending on its energy
K — thick plastic filter which attenuates low
energy photon radiations and absorbs all but the
highest energy beta radiation;
A — aluminium filter used with area K to assess
doses from photons with energies from 15 to 65
keV;
C — composite of cadmium and lead filters to
assess doses from thermal neutrons which
interact with the cadmium;
T — composite of tin and lead filters used with area
C to assess doses from thermal neutrons;
E — edge shielding to prevent low energy photons
entering around area T;
I — indium foil sometimes included to detect fast
neutrons

Advantages of Film Badges
• The holder creates a distinctive pattern on the film indicating
the type and energy of radiation to which it was exposed
(discrimination).
• Cumulative doses from beta, X, gamma and thermal
neutron radiations are calculated by measuring the optical
densities of film under the filters and comparing the results
with calibration films that have been exposed to known
doses.
• Films provide a permanent record that can, if necessary,
be reassessed.
• Pattern Discrimination:A single exposure tends to leave
sharp shadows on the film from the filters, whereas multiple
small exposures at different angles will leave a rim of
blurring around the filters

Disadvantages of Film Badges
• Holders must not be used if filters are missing or they
become contaminated.
• Radioactive contamination produces non-uniform black
patches on the developed film.
• Films are also adversely affected by light (if the wrapper is
damaged), heat, liquids, partial shielding and static
electricity discharges.
• The latent image on undeveloped film fades with time,
limiting possible wearing periods to three months in ideal
conditions – In India the wearing time is usually 1 month.
• Excessive fading occurs in relative humidity > 55%.

Thermoluminiscent Dosimeters
• Thermoluminescence 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°C.
• The intensity of the light may be
measured and related to the
radiation dose of the phosphor.

TLD Badges
• TLD chips available in India are
made from CaSO4.
• The doping material used is
13.3 mm
Dysporsium
0.8 mm
• Advantages of CaSO4:
12 mm – Useful over a large dose range
– Cheap and easily available
Cu/Al
– Highly sensitive
• Mixed with Teflon to form a tablet
• Sealed inside a plastic leaflet to
Plastic
protect from moisture

Advantages of TLD badges
TLDs are less affected than film badges by fading and ambient
•
conditions.
They are more appropriate for wearing periods of up to three
•
months.
They are also more suitable, but not ideal, dosimeters for
•
mixtures of weakly penetrating (less than 500 keV) beta and
(less than 20 keV) photon radiations.
Different materials can be used for different radiation types.
•
Can be reused unlike film badges.
•
Generally cheaper to use and maintain.
•
Can be modified for monitoring in any situation like extremities.
•
Have a linear response to dose received and are relatively
•
energy independent.

Pocket Dosimeters (Direct Reading)
• Provide the wearer with an immediate reading of exposure.
• Contains a small ionization chamber with a vol = 2cc
A central anode is provided along with a quartz fibre coated
•
with metal.
• When the anode is positively charged it repels the quartz
fibre
• When ionization occurs the quartz fibre moves towards the
anode.
• The device is pointed to a light source to read the exposure

Pocket Dosimeters
• Advantages:
– Immediate Readings
– Reusable.
• Disadvantages:
– Limited range,
– No permanent record,
– Sensitive to mechanical
problems

Biological Dosimetry
• Small changes in the body that result from exposure to
radiation may be measured as indicators of such exposure.
• The techniques which have been studied include:
– Reduced hair diameter
– Sperm abnormalities
– Reduced peripheral blood cell count
• Indications of the whole body dose may be calculated by
measuring the decline of circulating white blood cells
(lymphocytes) in the few days after an acute irradiation of
more than about 0.5 Sv.
• Checking for chromosomal aberrations is however the most
sensitive technique.

Chromosomal Abberations
• Lymphocytes are made to divide
and the number of dicentric
chormosomes are assessed.
• The incidence of dicentrics per
cell is dose dependent and
measurable above doses of:
– 100 mSv of Co gamma rays
60
– 50 mSv of 250 kVp X rays
– 10 mSv of approximately 1 MeV
neutrons.

Conclusions
Radiation protection is essential as no exposure to radiation can
•
be quantified as safe.
Three basic principles of Radiation protection include:
•
– Justifiable Use
– ALARA
– Dose limits
Three main methods of radiation protection are:
•
– Reduce time exposed to the source
– Increase distance from the source
– Use appropiate shielding methods
The responsibilty of radiation protection lies with the liscencee
•
and is overseen by authorised government agencies.
Constant vigilance and QA are essential in limiting unwanted
•
exposure to radiation

X X/2 5X
Thank You 50⁰
60⁰