Research A rticle
MCNP-X Monte Carlo Code Application for
Mass Attenuation Coefficients of Concrete at Different Energies
by Modeling 3 × 3 Inch NaI(Tl) Detector and Comparison with
XCOM and Monte Carlo Data
Huseyin Ozan Tekin
Vocational School of Heal th Service, Radiotherapy Department, Uskudar University, 34672 Ista nbul, Turkey
Correspondence should be addressed to Huseyin Ozan Tekin; huseyinozan.tekin@uskudar.edu.tr
Received May ; Revis ed June ; Accepted July
Academic Editor: Keith E. Holbert
Copyright © Huseyin Ozan Tekin. is is an open access article distributed under the Creative Commons Attribution License,
which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.
Gamma-ray measurements in various research elds require ecient detectors. One of these research elds is mass attenuation
coecients of dierent materials. Apart from experimental studies, the Monte Carlo (MC) method has become one of the most
popular tools in detector studies. An NaI(Tl) detector has b een modeled, and, for a validation study of the modeled NaI(Tl) detector,
theabsoluteeciencyof× inch cylindrical NaI(Tl) detector has b een calculated by using the general purpos e Monte Carlo
code MCNP-X (version ..) and compared with previous studies in literature in the range of – keV. In the present work,
the applicability of MCNP-X Monte Carlo code for mass attenuation of concrete sample material as building material at photon
energies . keV, keV, keV, . keV, . keV, and . keV has been tested by using validated NaI(Tl) detector. e mass
attenuation coecients of concrete sample have been calculated. e calculated results agreed well with experimental and some
other theoretical results. e results specify that this process can be followed to determine the data on the attenuation of gamma-
rays with other required energies in other materials or in new complex materials. It can be concluded that data from Monte Carlo
is a strong tool not only for eciency studies but also for mass attenuation coecients calculations.
1. Introduction
Since radioactive sources have been extensively used in many
elds such as medicine, industry, and energy applications,
radiation detectors have played a major role in detect i on
and quantication of ionizing radiations as well as radia-
tion protection procedures. Among all types of radiation
detectors, because of their higher detection eciencies of
gamma-rays, scintillation-based NaI(Tl) detectors have been
used widely for radiation spectroscopy and radioisotope-
based applica tion s including medical and industrial areas
[]. Furthermore, these types of detectors have shown reli-
able results for low level radioactive source measurements
because of higher detection eciency and suitable features
for operation at room temperature. However, the accuracy of
measurement depends strongly on some detection properties
of scintillation detectors such as detection eciency and
geometric eciency. Experimental gamma spectrometry is
an eective method to evaluate the eciency of NaI(Tl)
detectors. On the other hand, in recent years, the Monte
Carlo (MC) method has been widely utilized for eciency
determination studies [–]. Absolute eciency is dened as
the ratio of the number of counts produced by the detector
tothenumberofgamma-raysemittedbythesource(in
all directions). Some experimental and calculation studies
have reported on detection eciency determination [–].
Nowadays, radiation technology is starting to be used in a
variety of dierent areas such as nuclear power bases, particle
accelerators such as linac and synchrotron, and medical
facilities such as nuclear medicine and radiological facilities
and thus radiation protection becomes important. Shielding
is widely and eectively used method for protection from
radiation hazards; improvement of the shielding properties
of concrete which is a commonly used construction material
becomes important. ese studies can be performed using
Hindawi Publishing Corporation
Science and Technology of Nuclear Installations
Volume 2016, Article ID 6547318, 7 pages
http://dx.doi.org/10.1155/2016/6547318
Science and Technology of Nuclear Installations
ecient detectors. is study presents the use of NaI(Tl)
detectors modeled by Monte Carlo method during mass
attenuation coecients calculations. e validity of the mod-
eling geometry in Monte Carlo studies is critically important.
Based on this reason, rst step of this study was to create
a geometry by using MCNP-X and aerwards a validity
check to compare the eciency results of the detector with
available previous works in literature. e va lidated detector
is used for mass attenuation coecients calculations of
simple concrete sample at . keV (
241
Am), . keV (
133
Ba),
. ke V (
99m
Tc), . keV (
133
Ba), . keV (
137
Cs), .
(
60
Co), and . keV (
60
Co) photon energies which are the
most commonly used isotopes in n uclear investigations.
2. Material and Methods
2.1. Validation of Monte Carlo Model: Eciency Calculations
of Modeled NaI(Tl) Detector. Absolute eciency must be
known especially in radioactivity measurements. In t his
study, absolute eciency of modeled detector in a wide
energy range is calculated. e denition of the absolute
eciency is shown in
𝜀
abs
=
N
c
N
s
.
()
Here in (), N
c
denes the number of counts recorded by
detector and N
s
denes the number of radiations emitted by
the source (all directions such as isotropic source). In this
study, MCNP-X (Monte Carlo N-Particle eXtended) version
.. has been used for geometry design and calculations.
MCNP-X is a general purpose radiation transport code for
modeling the interaction of radiation with materials. MCNP-
X is fully three-dimensional and it utilizes extended nuclear
cross section libraries and uses physics models for p article
types. MCNP-X is a suitable and strong code that has a
capability for various studies. e MCNP-X capability of
detecting eciency and using dierent experimental and
Monte Carlo studies has been studied by Akkurt et al. [].
Also, using conditions of MCNP-X for dose dist ribution has
been studied by Tekin and Kara []. e detector geometry
was modeled according to Figure . And physical parameters
of related detector shown in Figure were taken from
Canberra Company and the same detector parameters also
used earlier in some experimental studies []. e detector
response function was determined by means of pulse-height
tally named F tally in the MCNP-X input le. is tally
scores the energy distribution of pulses created in a detector
by radiation. e net response is the spectra of pulses with
heights proportional to the frequency of events in distinct
energy bins.
All compounds and pure materials of detector were
dened in the MC input le. ese materials were aluminum
with density of . g/cm
3
, MgO with density of . g/cm
3
,
NaI(Tl) with density of . g/cm
3
,andSiO
2
with density of
. g/cm
3
. A screenshot of the modeled NaI(Tl) detector
by using MCNP-X code is shown in Figure . Due to the
simulation process in code, the total simulation geometry
is seen in Figure , and, as it can be seen, there is one
Al
MgO
NaI
c = 8.09
b=7.99
a=7.62
SiO
2
t=3.0
z=7.62
y = 0.185
x = 0.05
F : Schematic representation of NaI(Tl) detector considered
in simulation.
Central axis
Source
d
Pb shield
NaI (Tl)
F : Considered source and detector location in simulation.
cylindrical × i nch NaI(Tl) detector of height in crystal
. cm and diameter . cm with a monoenergetic isotropic
point source. Also, the source and detector assembly were
shielded by lead blocks.
Simulation equipment such as detector and lead blocks
havebeendenedindatacellcardandsurfacecardsec-
tions of MCNP-X input by considering dierent variables
such as geometry, location, dimension, and density. e
gamma-ray sources also have been dened in data card
CEL, ERG, DIR, POS, and PAR. Each variable has dierent
abilities during simulation. In present study, our variables
commanded source cell, energy, direction, source position,
and particle type, respectively. On the other hand, one of the
important denitions is material specication by considering
atomic number, mass number, and density for pure elemental
materials and ato mic number, mass number, elemental mass
Science and Technology of Nuclear Installations
T : Total eciency values for × inch NaI(Tl) detector with a point source located 𝑑 = . cm away from the front surface of the
detector.
Energy (keV)
Total eciency (𝑑= , cm)
Present work Yalcin et al. [] Vegors Jr. et al. [] Nakamura [] Miller and Snow [] MCNP-X error rate
. . . . . .
. . . . . .
. . . . . .
0.24
0.29
0.34
0.39
0.44
0.49
0 500 1000 1500 2000 2500 3000
Total eciency
Energy (keV)
Present work
Yalcin et.al.
Vegors et al.
Nakamura
Miller and Snow
F : Total eciency for × inch NaI(Tl) detector (𝑑=
0.001cm).
fraction, and density for compounds or mixtures. By con-
sidering these variables, material denitions of concrete have
been done in simulation. When some energy 𝐸 is deposited
into the detector, accounting the corresponding channel of
the spectrum is recorded. Somehow, the gamma-rays spectra
obtained in the simulations are very dierent from the spectra
obtained with the detectors. us, the resolution calibration
in the user code must be accounted for obtaining the realistic
spect ra []. As we know, the absolute eciency of detector
is dependent on the source-detector distance. In this study,
simulations were repeated by considering distance of 𝑑=
0.001cm source-detector distance. e MC calculations for
total eciency were presented in Table , for 𝑑 = 0.001cm
distance between source and detector. e geometric center
of detector was considered for location of point source.
MC calculations were done by using Intel Core i CPU
. GHz computer hardware. Also the comparisons were
made between present study and other studies in these tables.
e results show that with an increase in photon energy,
the total eciency is reduced for all distances in certain
energy values. e reduced eciency rates were given in
Table.Inthisstudy,theabsolutedetectoreciencyofa
modeled NaI(Tl) detector for dierent energies of photons
and for dierent source-detector distances was calculated
by using Monte Carlo method with MCNP-X code. A good
agreement was observed between the Monte Carlo and pre-
vious studies for 𝑑 = 0.001cm detector-source distance. Also
in the simulation, isotropic point sources were considered
for calculations. Total eciency values were calculated for
× inch NaI(Tl) detector for source-detector distances of
𝑑 = 0.001cm. Since the Monte Carlo method has become an
important tool in eciency studies, MCNP-X co de was used
for this study successfully.
AsweseefromFigure,goodagreementwasachieved
between eciency values. Somehow, Hybrid Monte Carlo
program which has been developed by Yalcin et al. []
requires rather short computing time. In the present simula-
tion, the run time was longer than in the Hybrid Monte Carlo
method. In this study, good agreement is achieved between
gamma energy and detector eciency. As the gamma energy
increases in interaction, the total detector eciency decreases
giventhatthepossibilityofaphotonbeingabsorbedinsideof
the detector decreases.
2.2. Mass Attenuation Coecients Calculations. To avoid
population and sta exposure to ionization radiation, work-
ing and public areas should be shielded. e main investiga-
tion required for such studies is mass attenuation coecient
values of building materials. Mass attenuation coecient
measures the probability of interaction of photon with the
material. Modeling the photon attenuation through materials
in a simulation environment gives more exibility and sim-
plicity of use and change of parameters instead of performing
an experimental study of mass attenuation coecients of
dierent materials. us, modeling of validated detector
geometry would be useful for future studies where the energy
valueshouldbechanged.Inthisstudy,wecarriedoutan
investigation on availability of MCNP-X Monte Carlo code
for calculation of mass attenuation coecients of simple
dened concrete used for mass attenuation coecients calcu-
lations [] e mass attenuation coecient is one of the most
important parameters for characterizing the penetration and
diusion of gamma-rays in any objective material [].
Mass attenuation coecients of investigated materials are
determined by the transmission method according to Beer-
Lambert’s law:
𝜇
𝑚
⋅𝑥=ln
(
𝐼
𝑜
𝐼
)
,
()
where 𝐼
𝑜
and 𝐼 are the incident and attenuated photon
intensity, respectively, 𝜇
𝑚
(cm
2
⋅g
−1
) is the mass attenuation
coecient, and 𝑥 is the thickness of the slab. In recent
years, many researchers have studied determination of mass
attenuation coecients theoretically and experimentally for
Science and Technology of Nuclear Installations
T : Concrete sample parameters (density = . g⋅cm
−3
).
Element Mass fraction (%)
H
C .
O.
Na .
Mg .
Al .
Si .
K.
Ca .
Fe .
various materials, such as some exp erimental studies per-
formed by Akkurt and El-Khayatt []. Eect of the boron
carbide aluminum metal matrix composite on radiation
shielding has been studied by Akkas¸ et al. []. Investigation
of biological materials and their attenuations by comparing
Monte Carlo and XCOM have been studied Medhat et al.
[]. Mass attenuation coecients of composite material
comparison have been studied by Medhat and Singh by using
Geant and XCOM []. In this study, e mass a tten uation
coecients of concrete sample dened in the simulation
packagewereobtainedbyusingtheMCNP-XMonteCarlo
code at ., ., ., ., ., ., and . keV
photon energies. Simple concrete material [] content ratios
have been dened as sample material in MCNP-X input.
Elemental structure and mass fractions of used concrete are
given in Table .
A complex elemental concrete has been dened in
MCNP-X due to elemental structure and mass fraction in
sample. However, the source was approximated as a point
source with other equipment in simulation such as concrete
sample and NaI(Tl) detector in lead (Pb) shielding material.
For a good interaction w ith NaI(Tl) detector, photons have
been collimated onto detector the window in simulation.
SinceMCNP-Xobtainstheprimarysourcesofnuclear
data, evaluations from the evaluated nuclear data le (endf)
system, evaluated nuclear data library (endl), and evaluated
photon data library (epdl) are highly capable of photonic
calculations. In this study, also some variance reduction
techniques have been applied such as cutting o energy and
reducing the types of observed particles in interaction such
as ignoring of electrons in mother world and equipment
geometries.
MC calculations were done by using Intel Core i CPU
.GHzcomputerandforonemillionstartingparticles
per run (NPS). Depending on the long calculation run t i me,
statistical error that has been obtained was less than %. Of
course, this error reduction not only depends on long run
time but also depends on variance reduction methods such
as cut-o energy applications in MCNP-X data card, ignoring
theunusedparticlesinsimulationsuchasneutronand
electron and optimized mother world volume in simulation
geometry.eaveragecelluxtally(F)hasbeenused
during mass attenuation coecient calculation. is type of
Source
Sample
Pb shield
NaI (Tl)
F : Schematic representation of simulation and locations of
modeled equipment.
tally makes use of what may be called a variance reduction
technique, namely, use of the next event estimator. F or each
source particle and each collision event, a deterministic
estimation is made of the uence contribution at the detector
point which is also shown in Figure . Since MCNP-X has
special material denition process, the user has to consider
the elemental composition of concrete and mass fractions
on the base of their chemical composition and weight
rates in denition of material in MCNP -X simulation. e
percentages by weight of the dierent elements for dierent
types of concrete are also given in Table .
3. Results
By considering the concrete sample which is dened in
MCNP-X code, mass attenuation coecient was calculated
in range of . keV, . keV, . keV, . keV, . keV,
. keV, and . keV photon energies. To observe the
transmissions of photons, dierent thicknesses of concrete
sample were used. In this study, source has been considered as
collimated isotropic source same as exp erimental conditions.
To obtain accuracy of results, energy spectra at detector face
quantied for each incident energy to determine amount of
photon downscattering within the sample for each energy. In
thisstudy,asimplemodeledgeometrywasusedtoestimate
the transmission of photons through the modeled concrete
sample with the dierent thicknesses.
Some comparison studies between Mo n te Carlo and
XCOM data have been performed by Demir et al. by using
FLUKA code. Table shows the calculated mass attenuation
coecients of concrete sample and photon energies by giving
calculated values by the XCOM [] database and previously
reportedMonteCarlovalues.Figureshowsthecalculated
mass attenuation coecients of concrete sample by MCNP-
X. Deviations (𝐷=𝐸
𝑎
−𝐸
𝑏
/𝐸
𝑏
× 100%) between this study
Science and Technology of Nuclear Installations
T : Mass attenuation coecients for the concrete sample.
Energy (keV) Present work (MCNP-X) Demir et al. [] XCOM Deviation (𝐷=𝐸
𝑎
−𝐸
𝑏
/𝐸
𝑏
×100%)
. . . . −. to +.
. . . . −. to .
. . . . −. to .
. . . . −. to −.
. . . . −. to .
. . . . −. to
. . . . −. to .
0
0.05
0.1
0.15
0.2
0.25
0 200 400 600 800 1000 1200 1400
Energy (keV)
Mass attenuation coecient (cm
2
/g)
F : Calculated mass attenuation coecients of concrete
sample by MCNP-X.
0
0.05
0.1
0.15
0.2
0.25
0 200 400 600 800 1000 1200 1400
Energy (keV)
XCOM
FLUKA
MCNP-X
Mass attenuation coecient (cm
2
/g)
F : Comparison of mass attenuation coecients of concrete
sample.
and other results from Demir et al. and XCOM are given in
Table .
e MCNP-X simulation code was employed to calculate
the values of the mass attenuation coecients for concrete
sample. e MCNP-X versus XCOM and previously cal-
culated values are plotted in Figure . It was found that
the simulated results of mass attenuation coecient values
of the composites for seven gamma-ray energies were in
good agreement with other data. Dierences between the
MCNP-X results of the XCOM and the MCNP mass atten-
uation coecients could be due to deviations from narrow-
beam geometry in the source-detector setup. Additionally,
dierences between the MCNP-X output results and FLUKA
output results could be due to dierent cross section and
data libraries of two dierent Monte Carlo codes. It has been
found that the MCNP-X simulation and modeled NaI(Tl)
detector can be applied to estimate the mass attenuation
coecients for various attenuator and energies in dierent
future studies. In addition, it can also be concluded that the
MCNP simulation code is a powerful method for evaluation
of photon interaction parameters of the dierent types of
materials.
4. Conclusion
In this study, as a validation of modeled detector, eciency for
dierent energies was obtained using MCNP-X code. In e-
ciency calculation of modeled NaI(Tl) detector, it was found
that obtained results were moderately similar to available
experimental data. Our MC model was capable of reproduc-
ing and conrming previous results on a commercial scintil-
lation detector. Our study showed that MCNP-X code results
are not only very similar to other experimental and Monte
Carlo results in eciency calculations but also very similar
to other XCOM and FLUKA Monte Carlo results in mass
attenuation coecients calculations. Dierences between the
results in mass attenuation coecient calculations could
be due to dierent cross section data between MCNP-X
and FLUKA and also could be due to computing time and
statistical error rates. So the same geometry and model could
be used for other applications such as attenuation studies with
dierentcompoundmaterialsaswellasgammaspectroscopy
for material characterization. As a conclusion topic, it can
be also concluded that these results are useful in developing
a better understanding of detector design by using Monte
Carlo method and the Monte Carlo method is used in mass
attenuation coecients calculation. ese results conclude
that MCNP-X Monte Carlo simulation is in well compatibility
with not only experimental data but also other Monte Carlo
codessuchasFLUKAcodeandcanbeappliedtopredict
the mass attenuation coecients for dierent attenuator and
energies and can be an alternative method for experimental