Tsoulfanidis N. Measurement and detection of radiation
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578
MEASUREMENT
AND
DETECIlON
OF
RADIATION
\
I-
Original surface
-
Surface removed
by
etching
/-
Track formed
by
etching
\
1-
Original track
Figure
16.11 Enlargement of a track by etching.
material such as mica or lexan.+ When the SSTR is exposed to neutrons, fissions
take place, and the fission fragments are imbedded into the SSTR material and
produce the tracks. For absolute measurement of fission rate, the observer
should be certain that all the fission fragments enter the SSTR and create
visible tracks; or, if some fragments are lost, a correction should be applied to
the observed fission rate. Some fission fragments will not escape no matter how
thin the fissionable deposit is (Fig. 16.12). As Fig. 16.12 shows,
a
fission fragment
emitted toward the SSTR will not reach it unless it is emitted within a cone
defined by the angle 28
=
2arccos(x/R), where
x
is the depth at which the
fission took place and
R
is the range of that fragment. If t
<
R,
where
t
is the
thickness of the deposit, the fraction of fragments reaching the SSTR is equal to
1
-
t/2R. If one places a second SSTR on the other side of the deposit, the
fraction of escaping fission fragments is doubled.
As
the thickness t decreases, a
greater fraction of fragments escape, but at the same time a smaller number of
fissions takes place and a smaller number of tracks is formed.
In addition to fission-fragment track measurements, thermal neutrons can
also
be
detected
by
alpha tracks from
(n,
a)
reactions with 6~i or
'OB.
Finally,
exa an
is a polymer developed by General Electric, with atomic composition C,,H,,O,.
Deposit of
fissionable
t
<
range of fission fragments
material
I
Figure
16.12
A
fission fragment emitted at an angle greater than
8
(8
=
arccos
x/R)
will
not reach
SSTR.
HEALTH
PHYSICS
FUNDAMENTALS
579
SSTRs have been used as monitors of alphas
particle^,'^
especially alphas
emitted
by
radon and its daughters.
16.9.4
The Bonner Sphere (the Rem Ball)
The Bonner sphere,29 named after one of the first people to study its features
and use it, is a neutron detector. It consists
of
a polyethylene sphere,
at
the
center of which a neutron detector is placed (Lil scintillator or BF, or 3~e
counter). With any one of these materials, the neutrons are detected through
the reactions
The polyethylene serves as the moderating material. The Bonner sphere has
been found to be very useful for neutron dose measurements because the
response of a 0.25- to 0.30-m (10-12 in) diameter sphere has an energy
dependence very close to the dose equivalent delivered by neutrons (Figs. 16.13
and 16.14). The line indicated as RPG in these two figures is the dose rate
H
per unit neutron
flux.
The similarity between detector response and dose
equivalent
H
is just a coincidence. This coincidence is utilized, in practice, for
the determination of neutron dose rate
H
in a neutron field of unknown energy.
Note from Fig. 16.13 that the sensitivity (efficiency) of the Bonner sphere is high
for high-energy neutrons (with high
Q
value) and lower for low-energy neutrons
(low
Q
value). For this reason, the number of counts recorded by the sphere
Figure
16.13
Sensitivity of a 10-in
(0.254
m) diameter Bonner sphere
surrounding
a
4
mm
X
4
mm
Lil
scintillator. For com~arison. the in-
I
1
1
verse of the response function RPG is
100
keV
1.0
MeV
10
MeV
also shown. The RPG gives dose rate,
neutron energy
H,
per unit neutron
flux
(Ref. 30).
580
MEASUREMENT AND DETECTION OF RADIATION
placed in an unknown spectrum 4(E) automatically includes a weighting factor
for all energies.
The match between response and neutron dose rate is not perfect, as Fig.
16.14 clearly shows, particularly for energy less than 100 keV. If the unknown
spectrum contains many neutrons in the energy range of discrepancy between
the two curves, then the dose rate will be overestimated for the following
reason. Consider
E
=
10 keV. The calculated response is about 1.7; the RPG is
about 0.3. The inverse of RPG is about 3.3, which means that the detector will
record 3.3 when the response ought to be close 1.7; hence, dose rate is
overestimated. Another point to mention is the underesponse at E
>
10 MeV
(Fig. 16.13); luckily, this energy is not important for neutron fields encountered
at nuclear power plants.
Another advantage of the Bonner sphere, in addition to its convenient
response function, is its complete insensitivity to gammas. This is the result of
relying for neutron detection on charged-particle reactions with high
Q
value,
thus making possible the complete rejection of pulses due to gammas with the
use of a proper discriminator level.
In the nuclear industry the Bonner sphere is known as the rem ball or rem
meter.
16.9.5
The Neutron Bubble Detector
The neutron bubble detector (trade name BD-100R) is a reusable, passive
integrating dosimeter that allows instant, visible detection of neutron dose. The
bubble detector consists of a glass tube filled with thousands of superheated
liquid drops in a stabilizing matrix. When exposed to neutrons, these droplets
vaporize, forming visible permanent bubbles in an elastic polymer. The total
number of bubbles formed is proportional to the neutron dose equivalent
H.
The bubbles can be counted manually or by a machine. Figure 16.15 shows the
response of the bubble detector as a function of neutron energy.
The bubble dosimeter is reusable; it is insensitive to gammas (the formation
of the bubbles is based on the stopping power of the recoil nuclei produced by
collisions with neutrons); it responds to $neutron energy range from 200 keV to
calculated
response
\
4
0.5
/-
measured
8c
Sj
0.2
-
Inverse of
RPG
curve
0.1
Thermal0.1eV l.0eV lOeV lOOeV 1keV lOkeV lOOkeV lMeV 1OMeV
neutron energy
Figure
16.14
Calculated sensitivity of a 10-in (0.254 m) diameter Bonner sphere compared with the
measured response (Ref. 30).
HEALTH
PHYSICS FUNDAMENTALS
581
Figure
16.15
The response of the neutron bubble dosimeter as
a
function of neutron energy
(Siemens,
1993).
about
14
MeV. The dose range extends from less than
1
mrem to
1
rem; its
useful life is
3
months, if recycled; its shelf life is
1
year.
16.9.6
The Electronic Personal Dosimeter
The electronic personal dosimeter (EPD) was developed by England's National
Radiological Protection Board and is now marketed by Siemen~.~' The EPD,
having the size and weight of a small pocket pager, uses the latest in integrated
circuitry technology. It is a solid-state device based on silicon diodes. Complete
details of the design are proprietary.
The objective of the EPD project was to produce a dosimeter that is
accurate over a wide energy and dose range, rugged, and can communicate with
a computer for data storage and subsequent analysis. All indications are that the
EPD satisfies these requirements. It is small and rugged and the energy range is
from 20
keV to
7
MeV for gammas and 250 keV to 1.5 MeV (average energy)
for betas. In terms of dose, the dosimeter can display doses from
1
pSv to
1
Sv
(0.1 mrem to 100 rem) and dose rates from
1
pSv/h to 10 Sv/h and is equipped
with audible and visual alarms that can be set only by authorized persons.
The EPD uses infrared links to interface with a computer, thus providing
the data to authorized persons who can read the dose received, add up previous
exposures, and establish dose-alarm threshold settings.
A
custom lithium battery
that lasts for a year powers the EPD. Each unit will be issued to a radiation
worker for a full year.
582
MEASUREMENT
AND
DETECTION
OF
RADIATION
16.9.7 Foil Activation Used for Neutron Dosimetry
Neutron dosimetry by foil activation is not used so much to record doses
received by personnel as it is to record doses to materials, instruments, or other
components that may suffer radiation damage as a result of neutron bombard-
ment. The principle of this method was presented in Secs. 14.4 and 14.6. A
target, in the form of a thin small foil, is exposed to the neutron field and
becomes radioactive. The relationship between activity and neutron
flux
is
where
N
=
number of targets
u
=
neutron absorption cross section
4
=
neutron
flux
A
=
decay constant of the radioisotope produced
After irradiation, the activity A(t) is counted and the
flux
is determined from
Eq. 16.34. Depending on the foil used (reaction involved), information about the
neutron energy spectrum may also be obtained. Information about the neutron
spectrum
+(E)
is necessary for the determination of the neutron dose equiva-
lent
H.
16.10 PROPER USE OF RADIATION
Since radiation may be hazardous, it is important that individuals who handle
ionizing radiation follow certain rules to avoid accidents. The official rules to be
followed by all persons licensed to handle radioactive materials have been
studied and proposed by such bodies as the ICRP and the National Research
Council (NRC), which is an arm of the National Academy of Sciences. The
proposed standards are adopted and enforced by federal agencies such as the
U.S. NRC and the Environment Protection Agency (EPA). The NRC and EPA
standards for protection against radiation are contained in
Code
of
Federal
~egulations.~~
The exposure limits, based on these guidelines, were discussed in
Sec. 16.8.
To protect personnel, areas where radiation sources are used are marked
with certain signs. The definitions of "radiation areas" and the corresponding
signs are as follows."
"Restricted area" means any area access to which is controlled by the licensee for purposes of
protection of individuals from exposure to radiation and radioactive materials. "Restricted
area" shall not include any areas used as residential quarters, although a separate room or
rooms in a residential building may be set apart as a restricted area.
"Radiation area" means an area, accessible to individuals, in which radiation levels could result
in an individual's receiving a dose equivalent in excess of 0.050
mSv (5 mrem) in
1
hour at 0.30
m from the radiation source or from any surface that the radiation penetrates.
HEALTH
PHYSICS
FUNDAMENTALS
583
"High radiation area" means an area, accessible to individuals, in which radiation levels could
result in an individual's receiving a dose equivalent in excess of
1
mSv
(100
mrem) in
1
hour at
0.30
m from the radiation source or from any surface that the radiation penetrates.
Radiation areas should be marked with the radiation symbol shown in Fig.
16.16
and with cautionary signs. If necessary, the radiation area should be roped
off or, if it is a room, should be locked to keep people out.
People who work in radiation areas or use radioisotopes should keep in
mind the following simple principle:
Radiation exposure should be avoided, if at all possible. If exposure is necessary, the risk from
the exposure should be balanced against the expected benefit. The exposure is justified if the
benefit outweighs the risk.
If exposure is justified, the employer and the employee should obey the
ALARA
principle (Sec.
16.8).
Satisfying
ALARA
means not just keeping exposure below
the maximum allowed limits; it means taking all possible "reasonably achiev-
able" measures to minimize the dose received
in
any task that has to be
performed. The best assurance against violating
ALARA
is constant education
and training of the radiation workers. Given below are some commonsense rules
that have proved helpful in reducing exposure.
1.
Try to avoid internal exposure. Substances enter the body by mouth (eating,
drinking), by breathing, through wounds, and by injection. Therefore, in
places where radioactive materials are handled, do not eat, do not drink, and
Figure
16.16
The standard radiation
symbol with dimensions as shown
has a yellow background, with the
hatched area being magenta or pur-
ple.
584
MEASUREMENT
AND
DETECTION
OF
RADIATION
cover all wounds. If the air is contaminated, wear a mask. Hands should be
washed after the operation is over, especially if no protective gloves were
used.
Stay close to the source of radiation for as short a time interval as possible.
Use protective covers, if this is the suggestion of the health physicist.
Place the source behind a shield or in a proper container.
If practical, wait for the radiation to decay to a safer level before handling it.
The exponential decay law is a helpful ally.
Stay as far away from the source as practical. The
flux
of a point isotropic
source decreases as
l/r2
(r
=
distance away from the source).
The shielding medium that should be used is not the same for all types of
radiation. Here are simple suggestions for the three types of radiation consid-
ered in this book.
Charged particles.
Charged particles have a definite range. Therefore, to stop
them completely, a shielding material with thickness at least equal to the range
(in that material) of the most penetrating particle should be placed between the
source and the worker.
A
few millimeters of metal will definitely stop all
charged particles emitted by radioisotopic sources. Some bremsstrahlung may
get through, though.
Gammas.
A
beam of gammas going through a material of thickness
t
is
attenuated by a factor exp(-pt), where
p
is the total linear attenuation
coefficient of the gamma in that medium. The higher the value of
p
is, the
smaller the thickness
t
that reduces the intensity of the beam by a desired
factor. In theory, the beam cannot be attenuated to zero level. In practice, the
attenuation is considered complete if the radiation level equals the background.
The attenuation coefficient
p
increases with the atomic number of the
material. The most useful practical element for gamma shielding is lead
(2
=
82).
Lead is relatively inexpensive, and it is easy to melt it and make shields with it
having the desired shape, size, and thickness.
Neutrons.
Shielding against neutrons is more difficult than shielding either
against charged particles or photons. If the source emits fast neutrons, the first
step is to provide a material that will thermalize the neutrons. Such materials
are water, wax, or paraffin.
Thermalized neutrons are easily absorbed by many isotopes. Examples are
115
113
In,
Cd, and
'OB.
Of these, the most practical to use
is
boron. It can be used
in powder form or be dissolved in water or liquid wax.
A
very simple but
effective shield for a source of fast neutrons is
0.15-0.30
m of wax or paraffin to
which boron has been added. The thickness of this borated material may
change, depending on the strength of the source and the amount of boron
HEALTH PHYSICS FUNDAMENTALS
585
added. Cadmium is very useful in sheet form.
A
cadmium sheet
3-6
mm
(d-
in) will stop a thermal neutron beam almost completely. Finally, shields have
been manufactured that are flexible, like rubber, yet are excellent neutron
atten~ators.~~
PROBLEMS
16.1
What is the total dose received by an individual standing in front of the open beam port of a
research reactor for
10
s under the radiation levels listed in the figure below?
Core
C-
I
10'
fast
neutronsl(ma s)
lo9
thermal
neutrons/(ma s)
lo7
yl(m's) (E,
=
2
MeV)
10'
yl(ma
S)
(Ey
=
0.5
MeV)
lO1O
pl(ma
s)
(ED,
,,,
=
1.2
Mev)
16.2
What is the dose rate per curie of 24Na if it is shielded by 0.025 m of lead as shown below?
24
Na emits a 1.37-MeV gamma and a 2.75-MeV gamma 100 percent of the time, betas with
Em,,
=
4.17 MeV 0.003 percent of the time, and betas with
Em,,
=
1.389 MeV 100 percent of the
time.
16.3
As
a result of carelessness, a worker inhaled 1 yg of "'~rn. Considering alpha particles only
and assuming that the americium is spread uniformly in the bones, (a) what is the dose rate at the
time of the accident and (b) what is the total dose to that individual? For 24'~m,
T,
=
433 years and
T,
=
200 years, mass of bones
=
10 kg.
16.4
What is the dose rate 0.30 m away from 1 Ci of 60~o if (a) the attenuating medium is air, and
(b) if the source is shielded by 0.01 m of aluminum?
16.5
What is the thickness of a lead container that will result in a dose rate at its surface of 2.5
586
MEASUREMENT AND DETECTION
OF
RADIATION
mrem/h
=
2.5
X
lo-'
Sv/h, if it is used to store
1
Ci of Iz4sb? lZ4sb emits the following gammas:
Energy Energy Energy
(MeV) Intensity
(%)
(MeV) Intensity
(%)
(MeV) Intensity
(%)
0.603
97 0.967 2.4 1.37 5
0.644
7 1
.048 2.4 1.45 2
0.720
14 1.31 3
1.692 50
2.089 7
16.6
What is the dose rate due to the lZ4sb of Prob.
16.5
at a distance equal to the thickness of the
container, if the attenuating medium is air?
16.7
A
1-Ci sample of 60~o is stored behind a concrete and lead shield as shown below. What is the
dose rate at point
P?
Concrete
16.8
The isotope 99m~c is used
in
uivo
for diagnostic purposes in humans. It emits X-rays with
energy
140
keV and betas with Em,,
=
0.119
MeV. If a person is injected with
1
pCi of this isotope,
what is the total dose to the brain, assuming that all the isotope is uniformly distributed there?
(Mass of the brain
=
1.5
kg.)
16.9
If all the water in the human body were suddenly changed to T20(T E~H), what would be the
total dose to that individual? Assume
61
percent of the body is water.
16.10
If
1
pCi of
239~~
is inhaled by breathing and gets into the lungs, what is going to be the total
dose to that individual?
(T,
=
24,000
years,
T,
=
200
years, mass of lungs
=
1
kg.)
16.11
Calculate the total dose rate at the center of a spherical submarine submerged in contami-
nated water with activity
1
Ci/m3 of I3'cs. The submarine is made of steel
0.025
m thick. Its radius
is
1.5
m and it is filled with air
at
1
atm.
16.12
What is the annual dose to a person due to the
40~
found in every human body? The isotopic
abundance of
40~
in potassium is
0.0119
percent. The human body contains
1.7
X
kg of
potassium per kg. 40K emits the following radiations:
Pa~tiele Energy (MeV) Intensity
(%)
16.13
A
"hot particle" (a tiny speck of radioactive material) was lodged on the palm of a radiation
worker, from where it was removed after
10
min of scrubbing. What is the estimated total dose
HEALTH
PHYSICS
FUNDAMENTALS
587
received by the tissue exposed to this radiation if it was determined that the radioactivity came from
a beta emitter with
Em,,
=
1.7 MeV, half-life equal to 2 days, and initial estimated activity equal to
1.3 mCi?
16.14
Show that the doubling dose for radiation-induced mutations is between 0.5 and 2.5 Sv
(50-250 rem) if the probability to induce a mutation by irradiation is 2-0.4 per Sv.
16.15
Prove that the probability that a fission fragment will escape from one side of a fissile deposit
is t(l
-
x/R), where
R
is the average range of fission fragments and x is the depth in the deposit
where the fission took place.
16.16
Prove that the total number of fission fragments escaping from one side of a fissile deposit of
thickness
t (t
<
range) is equal to
F(t
-
t2/2R), where
F
is the number of fissions per unit volume.
BIBLIOGRAPHY
Attix, F.
H.,
Introduction to Radiological Physics and Radiation Dosimetry,
Wiley, 1986.
Attix, F.
H.,
and Tochilin,
E.,
(eds.),
Radiation Dosimehy
I,
11, 111, Academic, New York 1969.
Cember,
H.,
Introduction to Health Physics,
Pergamon,
1969.
Fitzgerald,
J. J.,
Brownell,
G. L.,
and Mahoney,
F.
J.,
Mathematical Theory of Radiation Dosimetry,
Gordon and Breach, 1967.
Fleischer, R.
L.,
Price, P. B., and Walker, R. M.,
Nuclear Tracks in Solids,
University of California
Press, Berkeley, Calif., 1975.
Hine, G.
J.,
and Brownell,
G.,
Radiation Dosimetry,
Academic, New York, 1956.
Miller,
K.
L.,
Handbook of Management of Radiation Protection Programs,
2nd ed., CRC Press, Boca
Raton, Fla., 1992.
Morgan,
K.
Z.,
and Turner,
J.
E., (eds.),
Principles of Radiation Protection,
Wiley, 1967.
Prasad,
K.
N.,
Handbook
of
Radiobiology,
CRC Press, Boca Raton, Fla., 1984.
Shani,
G.,
Radiation Dosimetry Instrumentation and Methods,
CRC Press, Boca Raton, Fla., 1991.
REFERENCES
1. Dose Equivalent ICRU Report 19 (suppl.), Washington, D.C. (1973).
2.
1977 Recommendations of the International Commission on Radiological Protection
(ICRP Publica-
tion 26), Pergamon, 1977.
3. Report of the RBE Committee to ICRU,
Health Phys.
9:357 (1963).
4. Standards for Protection against Radiation,
Code of Federal Regulations,
Title 10, Part 20, Dec.
1992, chap. 1.
5.
Engineering Compendium on Radiation Shielding,
vol.
1:
Shielding Fundamentals and Methods,
Springer-Verlag, 1968.
6. Stamatelatos, M. G., and England,
T.
R.,
Nuclear Sci. Eng.
63:204 (1977).
7. Neutron and Gamma-ray Flux-to-Dose Rate Factors, American National Standard, ANSI/ANS-
6.1.1-1991.
8.
Report of the ICRP Committee
I1
on Permissible Dose for Internal Radiation (1959).
Health
Phys.
3:146 (June 1960).
9. The Effects on Populations of Exposure to Low Levels of Ionizing Radiation: 1980 (the
BEIR-111 report), National Academy Press, Washington, D.C., 1980.
10. Health Effects of Exposure to Low Levels of Ionizing Radiation (the BEIR-V report), National
Academy Press, Washington, D.C., 1990.
11.
1990 Recommendations of the International Commission on Radiological Protection
(ICRP Publica-
tion 60), Pergamon, 1991.
12. Dudley, R.
A.,
in F.
H.
Attix and W. C. Roesch (eds.),
Radiation Dosimetry,
11, Academic Press,
New York, 1966.