Radioactivity Measurements for Some Building Materials in Yemen and Simulation of the Annual Effective Dose

Abstract

Fifty three samples of building materials were collected from two governorates in Yemen (Taiz and Hodeidah); these materials are used mostly in Yemen. Samples were measured for gamma radiation using HPGe detector. The specific absorbed dose rates due to the three natural radionuclides 226 Ra, 232 Th and 40 K in the most used types of building materials in Yemen such as (ordinary concrete, granite stone and cement brick) were calculated. The calculations were done for a model of spherical shaped room of radius 150 cm, thickness 30cm and variable density that varies according to the supposed material in the two selected cities (Taiz and Hodeidah). Stranden model is considered here with some modification in order to fit the specifications of the room in Yemen. The calculated annual effective dose rates for the ordinary concrete in the two cities were 329.452 and 294.250 (Hodeidah and Taiz) µSv/y respectively, in the granite stone was 1029.829 µSv/y, and in the cement brick was 929.497 µSv/y. Index Terms—Activity concentrations, annual effective dose, MCNP code and Yemen radioactive contamination.

Full text

Abstract—Fifty three samples of building materials were
collected from two governorates in Yemen (Taiz and Hodeidah);
these materials are used mostly in Yemen. Samples were
measured for gamma radiation using HPGe detector. The
specific absorbed dose rates due to the three natural
radionuclides 226Ra,232Th and 40K in the most used types of
building materials in Yemen such as (ordinary concrete, granite
stone and cement brick) were calculated. The calculations were
done for a model of spherical shaped room of radius 150 cm,
thickness 30cm and variable density that varies according to the
supposed material in the two selected cities (Taiz and Hodeidah).
Stranden model is considered here with some modification in
order to fit the specifications of the room in Yemen. The
calculated annual effective dose rates for the ordinary concrete
in the two cities were 329.452 and 294.250 (Hodeidah and Taiz)
µSv/y respectively, in the granite stone was 1029.829 µSv/y, and
in the cement brick was 929.497 µSv/y.
Index Terms—Activity concentrations, annual effective dose,
MCNP code and Yemen radioactive contamination.
I. INTRODUCTION
The environment in Yemen is varied between plain, desert
and volcanic islands. These varieties imposed the citizens to
use the available building materials. In the areas with
mountains, the nature of land has imposed the Yemenis to use
rocks as the basic building material. Because of the big
varieties of building materials used in different cities around
the Republic of Yemen and the shortage of information about
the radioactivity of these materials, we did our research in
order to make a regulation of the building materials in Yemen.
Some very common building materials like granite stone and
Cement brick were found of high value of the activity
concentrations. A theoretical model was set for a Yemeni
room with approximate specification of the room in Yemen
to calculate the indoor exposure dose rate.
This modeled room was established using MCNP code
(Monte Carlo N-particle transport computer code) and some
mathematical treatments [1].
II. EXPERMINTAL WORK
Fifty three samples were collected randomly from two
governorates; Taiz and Hodeidah in Yemen. Taiz represents
the mountainous areas located between latitudes 14 ° and 12°
to the north of the equator, and between longitudes 45° and
43° to the east of Greenwich, However Hodeidah represents
the plain areas, it lies the west of red sea cost in the area
Manuscript received May 14, 2012; revised June 21, 2012.
The authors are with Physics Dept., Faculty of science, Cairo University,
12613Giza, Egypt (email: mmsherif@eun.eg; safayusuf5@gmail.com).
between latitudes 14° and 16° to the north of the equator, and
between longitudes 42° and 43° to the east of Greenwich.
Some of the raw building materials were collected from the
places where they sold and some were collected from their
original resources (mining places) or from actual building
sites. The artificial building materials were collected from the
places they were sold in either from shops or from factories
so that the collected samples covered the most citizens' use of
these materials.
III. SAMPLE PREPARATION
The collected samples were saved in plastic bags. Then
they were crushed by hummer, and sieved through 0.8 mm
mesh sieve. Each sample was weighted and stored in a sealed
marinelli beaker for more than four weeks to reach the
secular equilibrium between 226Ra and its short lived
products.
IV. EXPERIMENTAL RESULTS
We have used the HPGe detector to measure the activity
concentrations in our samples in order to assess the individual
exposure dose rate and to estimate the risks from spending
most of our lifetimes inside buildings. The activity
concentrations for raw materials were ranged between (1.858
±0.333-154.216±6.974),(0.288±0.165-229.141±3.398),(3.38
9± 0.266-1701.338±59.572) Bq/Kg for 226Ra, 232Th and 40K
respectively, whereas the activity concentrations for the
industrial materials were ranged between(0.209±0.155
-180.950±6.922),(0.491±0.088-252.854±3.939) and (2.480
±0.958-1017.220±12.080) Bq/Kg for 226Ra,232Th and 40K
respectively. The highest value of the activity concentrations
was in the cement brick (180.950±6.922 and 252.854±3.94)
Bq/Kg for 226Ra and 232Th respectively, and granite stone
(154.216±6.974 and 229.141±3.398) Bq/Kg for 226Ra and
232Th Radionuclides, the concentration of these radionuclides
increases the risks from these radionuclides inside buildings,
as these bricks are used as the main building materials in
most of the buildings around the country and especially in the
two areas under study in this research.
V.
ACTIVITY ANALYSIS
A. The Radium Equivalent
The radium equivalent Raeq was calculated for all samples.
Because the distribution of 238U,232Th and 40K in nature is not
uniform, the radium equivalent Raeq is proposed to
comparing the specific activity of material containing
different amount of 238U, 232Th and 40K, and it is defined as a
Radioactivity Measurements for Some Building Materials
in Yemen and Simulation of the Annual Effective Dose
M. M. Sherif and Safa.Y. Abdo
International Journal of Environmental Science and Development, Vol. 3, No. 4, August 2012
319

weighted sum of the activity concentrations of the 238U, 232Th
and 40K.The measured specific activity (Bq/Kg) of 238U, 232Th
and 40K for each sample are used to calculate Radium
equivalent Raeq using the following equation [2] :
( ) 1.43 ( ) 0.077 ( )
eq
Ra CRa CTh CK=+ + (1)
where C(Ra), C(Th) and C(K) are the activity concentrations
in Bq/Kg. For the limitation of the annual effective dose to be
1mSv for the population, the maximum value of this index
must be less than 370Bq/Kg.
B. The External Hazard and the Internal Hazards
For the safe use of the materials in the Yemeni buildings
and to limit the annual effective dose to be 1mSv for the
population the external hazard Hex and the internal hazard
indices Hin which are given by [2-3]:
1
370 259 4810
Ra Th K
ex CC C
H=++ ≤
(2)
And
1
185 259 4810
Ra Th K
ex CC C
H=++ ≤
(3)
Should be less than unity.
TABLE I: THE CALCULATED RADIUM EQUIVALENT eq
Ra , THE INTERNAL
HAZARD in
H AND THE EXTERNAL HAZARD ex
H IN INDUSTRIAL AND
RAW MATERIALS
Index Industrial materials
(Bq/Kg)
Raw materials (Bq/Kg)
Radium
equivalent
1.340±0.349–
603.698±13.463
5.718±1.747 -
593.177±12.856
Internal
hazard 0.004±0.001- 2.120±0.055 0.023±0.007 – 2.019±0.054
External
hazard 0.004±0.001-1.630±0.036 0.006±0.001-1.602±0.035
From the “Table. I” the values of Raeq , Hin and Hex are
almost twice world average value.
C. The Absorbed Dose
The absorbed dose rate in air (D) in nGy/h, resulting from
the natural specific activity concentration of 238U, 232Th and
40K in Bq/Kg, at a height of 1 m above the ground was
calculated according to this formula [4]:
1 226 232 40
( ) 0.429 0.666 0.42DnGyh Ra Th K
−=++
(4)
where the contribution from the 238U has been replaced with
the decay product 226Ra. For the industrial materials the
absorbed dose ranged [13.734±1.559-279.393±9.834] Gy/h
and for the raw material it ranged [2.60±0.62– 279.47±10.60]
Gy/h. Some samples has absorbed dose higher than the
estimated average global terrestrial radiation of the range
24-160 nGy/h [5]. It is clear that the absorbed dose in the
granite stone and cement brick collected from Taiz and
Hodeidah are (279.470±10.602 Gy/h) and (279.393 ±9.834
Gy/h) respectively, are higher than the calculated values for
some soil and stones samples collected from Juban town in
Yemen [6].
D. The Investigative Level
Another hazard index called the investigative level was
determined for all the samples according to [7]:
300 / 200 / 3000 /
Ra Th K
CC C
I
B
qKg BqKg BqKg
=++
(5)
The investigative level was ranged between [0.020±0.006
– 2.142±0.045] for the raw materials, and ranged between
[0.005±0.001–2.132±0.047] in the industrial material.
According to the European commission the activity
concentration shall not accede the following values
depending on the dose criterion, the way and the amount the
material used in a building [7]:
TABLE II: DOSE CRITERION RANGE
Dose criterion 0.3mSv a-1 1mSva-1
Materials used in bulk amount,e.g. concrete I≤ 0.5 I≤ 1
Superficial and other materials with
restricted use: tiles ,boarded I≤ 2 I≤ 6
VI. MODELING
Our model is a developed for Mustonen model [8],
although it is a modification of Stranded model [9]. The
indoor exposure dose rate at a point in dwelling is written as
following:
.
2
(,)
()()
4
di
D
ii i a i BEiskc
x
EN E E e dV
l
ρ
μ
−
=∑∫
Π (6)
where x. the exposure dose rate,
ρ
is the density of the
material, C is the activity per unit weight, k is the coefficient
to change the exposure in to Roentgen unit ;( k = 1.462 X10-2
R/Mev.cm3) , E is the photon energy, N(Ei) is the number of
photons with energy Ei emitted per unit primary
disintegration, µa(Ei) is the linear absorption coefficient in air
µm(Ei) is the attenuation coefficient in the material BD(Ei,s) is
the build-up factor, S is the distance the photon travels in the
material, L is the distance from the source point, V is the
volume of the room and di is the optical distance between the
source and the detection point which is given by:
()( )
mi a
di s E L s
μμ
=+−
(7)
The linear attenuation coefficients (µm) for the mostly used
building materials in the selected cities (ordinary concrete,
granite and cement brick) have been calculated, by using
MCNP code. Our summation in “(6)” is done over only 18
selected energy lines used in this simulation model to
calculate the exposure dose rate.
VII. CALCULATION OF THE LINEAR ATTENUATION
COEFFICIENTS
A theoretical model was built to calculate the linear
attenuation coefficients for the selected energy lines using
MCNP code [1]. MCNP code is a Monte Carlo N-particle
transport computer code created and developed by Los
Alamos National Laboratory that can be used for neutron,
International Journal of Environmental Science and Development, Vol. 3, No. 4, August 2012
320

photon, electron, or coupled neutron/photon/electron
transport. The selected geometry for the theoretical model
here is a sphere. We detected the flux resulting from the
radionuclides inside the spherical layer of the supposed
material that we want to calculate the linear attenuation
coefficient for it by a standard tally in MCNP code named
(F5).
This sphere has a thickness of 30 cm, where the detection
is performed using a ring detector of radius of 50 cm in centre
of the room in about 150 cm from the internal wall (F5 tally).
We used a matrix of group of the gamma energies of
238U,232Th and40K in each running of MCNP. We supposed
that the sphere layer built of concrete, and the initial flux ϕ0
was detected first inside a vacuum sphere, after that we filled
the sphere with the ordinary concrete, got the attenuated flux
ϕ resulting from existing of the concrete material. By using
the simple relation of the photon attenuation equation:
0
x
e
μ
φφ
−
= (8)
where x is the distance from the entire wall to the detection
ring applied by (F5) tally in the MCNP input, x=150.We got
the values of the linear attenuation coefficients µm “Table
III.”, and these values of concrete are in a good agreement
with the published values by Mustonen [8]. We repeated the
same steps to calculate the linear attenuation coefficients µm
of granite stone and cement brick. Each time we changed
only the chemical compositions of the studied material in the
input file and all their densities. The attenuation coefficients
for concrete, granite stone and cement brick listed also in
Table. III.
TABLE III: THE ATTENUATION COEFFICIENTS FOR THE ORDINARY CONCRETE, GRANITE AND CEMENT BRICK
E (Mev) µ concrete
(µ cm-1)
µ granite
(µ cm-1)
µ cement brick
(µ cm-1)
E (Mev) µ concrete
(µ cm-1)
µ granite
(µ cm-1)
µ cement brick
(µ cm-1)
0.063 0.370 0.580 0.081 0.351 0.240 0.320 0.150
0.092 0.280 0.350 0.071 0.583 0.180 0.230 0.070
0.186 0.230 0.260 0.078 0.609 0.190 0.250 0.120
0.209 0.195 0.260 0.0128 0.860 0.160 0.200 0.100
0.238 0.220 0.290 0.0120 0.911 0.170 0.210 0.120
0.277 0.220 0.280 0.075 0.968 0.170 0.210 0.110
0.295 0.220 0.270 0.123 1.120 0.160 0.170 0.090
0.300 0.240 0.290 0.158 1.464 0.140 0.170 0.090
0.338 0.210 0.260 0.107 1.760 0.140 0.160 0.090
VIII.
THE SPECIFIC EXPOSURE DOSE RATE
From “(6),” the specific exposure dose rate per unit
activity concentration (Q) is given by the following relation:
2
(,)
()()
4
di
D
ii i a i BEis
k
QENEE edV
l
ρ
μ
−
=∑∫
Π (9)
The supposed geometry of a spherical shaped modelled
room of 150 cm radius and wall thickness of 30cm is shown
in “ Fig. 1”, this is the easiest for modeling since the sphere is
one dimensional, and the spherical shape is compatible with
some houses in Yemen which have been domed the roof.
According to Koblinger [10], the good agreement of the data
from this approximation with those obtained for rectangular
shaped room shows that the shape of the room hardly affects
the dose rates, so for the simplicity we have chosen the
spherical shaped room. The area of windows and doors in our
supposed room was taken in to consideration however they
act as shields against gammas coming from terrestrial sources
or walls of other rooms. We use in our calculation the
Berger’s formula of the build-up factor has the simplicity of
the linear form but fits the buildup Factor data over a long
range, and it is given by [11].
(,)1 () ()exp( )
iimim
B
Es aE Es b s
μμ
=+ (10)
Our assumption is based on a spherical shaped room of 150
cm and thickness of 30 cm. These specifications are
compatible with the design of the Yemeni room. For the
calculation of the flux at a point in the centre of the room, or
any other point from a volume source like concrete, or granite
room we supposed that the gamma radiation source as a point
source inside the material wall “q”, then the contribution
from all point sources to the total flux is added in the integral
over the thickness of the room, while the integration was
done using the spherical coordinates.
2sindV r d d dr
θθ
φ
= (11)
“Fig.1”shows the geometry used in our calculations, from
this figure we can see that:
)12(rl = )13(
Qp rrr −=
)14(
0
rrs −=
Fig. 1. Modeled room
, is the distance from the internal wall of the spherical
room to the detection point (p) in the centre. By substituting
in “(9)” and the integrated part using MATHEMATICA 5.2
International Journal of Environmental Science and Development, Vol. 3, No. 4, August 2012
321

soft ware, we got the specific exposure dose rate inside the
sphere.
TABLE IV: SPECIFIC EXPOSURE DOSE RATE
The material The specific exposure rate in µRh-1 / Bq-1Kg-1
226Ra 232Th 40K
Concrete 0.046 0.096 0.008
Granite 0.043 0.087 0.008
Cement brick 0.044 0.099 0.005
IX. THE ANNUAL EFFECTIVE DOSE RATE IN THE MIDDLE OF
A ROOM BUILT WITH DIFFERENT TYPES OF BUILDING
MATERIALS
The annual effective dose rate is calculated according to
this relation [12]:
()
wall Ra Ra Th Th K K
E
YT C Q C Q C Q m=++
(15)
where Y is the factor that converts the absorbed dose in air to
effective dose in humans (Sv/Gy), T is the indoor occupancy
factor and CRa,CTh and CK are radioactivity concentrations for
226Ra, 232Th and40K, respectively. The quantities QRa, QTh and
QK are the respective specific absorbed dose rates which have
been calculated for a typical Yemeni room built with concrete,
granite stone and cement brick, and m is the fraction of the
wall is made up of the material type (concrete, granite or
cement brick), it is supposed here to be equal to (32%, 58%
and 51%) for concrete, granite and cement brick respectively.
TABLE V: ANUUAL EFFECTIVE DOSE RATE FOR DIFFERENT TYPES OF
BUILDING MATERIALS
City Type of
material
The total annual effective dose rate
of whole room (µSv/y)
Hodeidah Concrete 329.452
Taiz Concrete 294.250
Distributed around
Yemen Granite stone 1029.829
Distributed around
Yemen
Cement
brick 929.497
The annual effective dose for concrete in Hodeidah and
Taiz 329.452 and 294.250 µSv/y respectively, is less than
that in Jordan (470 µSv/y) [13] ,Nigeria (400 µSv/y) [12],
Cuba (429.2 µSv/y) [14] and less than the dose in typical
building in Hong Kong (1459 µSv/y) [15] but lied within this
range of the total (outdoors plus indoors) annual effective
dose equivalent from terrestrial gamma radiation, averaged
over the world’s population (30 µSv/y -400 µSv/y) [16].In
granite stone the annual effective dose(1029.829 µSv/y)is
higher than that obtained in Jordan (520 µSv/y) [13] ,but
within the range of the effective dose rate calculated for
granite stone in Iran (480-1050 µSv/y) [17],whereas it is
twice the world average range .The effective dose rate
calculated for cement brick in this work is higher than that
dose calculated in Jordan (442 µSv/y) [13] , Cuba (258.59
µSv/y)[14], and also is twice the world average range.
X.
CONCLUSION
We observed widespread use of building materials like
cement brick and granite with high values of the activity
concentrations of the three studied radionuclides and their
resulted absorbed dose. We intend to make guideline for
those responsible to make a regulation for the specifications
of the building materials in Yemen.
ACKNOWLEDGMENTS
The authors would like to thank the Institute for
Radioecology and radiation protection (ZSR), Hanover,
Germany and Egyptian atomic energy authority, for
providing some facilities in the measurements of the samples.
Also the authors would like to thank Dr. Shaban Harb in
South Valley University, faculty of science, Egypt and Dr.
Hanan Diab in Egyptian atomic energy authority for their
help.
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