Modelling of Paraffin Shielding for BNCT Facility at ...
Transcript of Modelling of Paraffin Shielding for BNCT Facility at ...
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Modelling of Paraffin Shielding for BNCT Facility at Kartini Reactor
Research using MCNPX Hana Alfiani Lutfin 1, Sukmaji 2, Widarto 3
1Department of Physics , Faculty of Methematics and Natural Science, Jendral Soedirman University, Purwokerto 53122 xCentre of Accelerator Science and Technology, National Nuvlear Agency of Indonesia (BATAN), Yogyakarta 55281
ARTICLE INFO A B S T R A C T
Article history:
Received: 16 August
Received in revised form: xx month 20
Accepted: xx month 20xx
Keywords:
Shielding Paraffin Lead
MCNPX
Dose Rate
The development of cancer in the world is very high.
According to the World Health Organization (WHO) 1.69
million people die from cancer. While the case of cancer in
Indonesia is also not much different. Areas that have high
pravalence namely D.I Yogyakarta. Cancer that has
become a scourge for many people, must be considered.
There are many treatments done such as chemotherapy,
surgery, and radiation. However, there is radiotherapy
using neutron capture on boron-10 with energy 0.025 Ev.
This treatment does not damage other tissues because the
resulting particles such as He-4 and Lithium-7 have ranges
that are at a distance of 4.5-10 μm so that the deposited
energy is limited to the distance of a single cell diameter.
The treatment is Boron Neutron Capture Therapy (BNCT).
There are several BNCT facilities such as reactor, radial
piercing beamport, thermal column, and shielding.
Function of shielding is absorption of neutron and alpha
radiation. Therefore, shielding is made using excellent
paraffin material in absorbing neutron radiation. In addition
to paraffin there are also other materials such as Lead as
paraffin casing. In shielding simulation result using
MCNPX software resulted dose rate of radiation exposure
outside BNCT facility in vitro in vivo test that is equal to
6.5 μSv / h. The thickness of shielding paraffin used is 40
cm, Pb casing 25 cm, and 5 cm soft tissue.
© 2020 IJPNA. All rights reserved.
1. INTRODUCTION1*
Cancer is an abnormal growth of tissue cells
in the body. Cancer is one of the cause of death in the
world. According to World Health Organization
(WHO) data in 2015, 8.8 million people died from
cancer. The most common causes of death from
cancer include lung cancer of 1.69 million people
died, liver cancer 78800 people died, colorectal
cancer 774 000 people died, stomach cancer 754 000
people died, and breast cancer 571000 people died
(WHO Media Center, 2017). Basic Health Research
Data 2013, Ministry of Health Research and
Development Agency and Target Population Data,
Pusdatin Ministry of Health of 347,792 suffered from
cancer in Indonesia. The highest number of cancer
patients in Indonesia according to Health Research
1* Corresponding author
E-mail address: [email protected]
Data RI that is the area of Central Java and East Java
with the number of patients is as many as 68,638 and
61,230 inhabitants. The highest prevalence of cancer
patients is D.I. Yogyakarta with 4.1% pravalensi
(Kementrian Kesehatan RI Pusat Data dan Informasi
Kesehatan, 2015).
Cancer treatment developed by the method of
cleavage, psychotherapy, immunotherapy, surgery,
and radiation. Diagnosis of cancer followed by
surgery, especially diagnosis after cancer screening
will cause over treatment and cause damage to body
tissues. (Benjamin, 2014). Therefore, radiotherapy
that provides low side effects is needed. BNCT is a
radiotherapy that uses thermal neutron capture on
boron-10 by a low-energy neutron with 0.025 eV
resulting in two high Linear Energy Transfer
particles. The two particles are alpha and lithium-
7(Bortolussi et al., 2018). These particles serve to kill
Indonesian Journal of Physics and Nuclear Applications Volume 5, Number 2, June 2020, p. 31-37
e-ISSN 2550-0570, © FSM UKSW Publication
32
targeted cancer cells, without damaging other tissues
(Lai & Sheu, 2017).These particles have a range that
is at a distance of 4.5-10 μm so that the energy is
deposited is limited to the distance of a single cell
diameter (Moss, 2014). The neutron loading reaction
can be seen in the following scheme:
10B+nth (0.025 eV) [11 B]
(Kageji et al., 2014)
Reaction of 93.7% to He (alpha particle) with
energy 1.47 MeV, Li with energy 0.84 MeV and 0.48
MeV gamma energy; and the rest is lithium decay
(6.3%), which produces alpha with Li and each has
an energy of 1.78 MeV and 1.01 MeV. The ionisation
particles are proven to be effective agents, have high
linear energy transfer (LET) in the range of 100 KeV
/ μm, and have a very high efficiency to maintain
(Payudan, Aziz, & Sardjono, 2016). BNCT principles
can be seen in Fig 1.
(Tsurayya, 2017)
The principle of BNCT is to use an epithermal
neutron with a range of 0.5 eV-10 keV (Shaaban &
Albarhoum, 2015). BNCT can be used in nuclear
facilities and hospitals that develop neutrons (Made,
Dwiputra, Harto, & Sardjono, 2016).
In Taiwan there are BNCT groups that are
building BNCT-based accelerators. It aims to know the
characterization of the radiation field and the shielding
requirements to be made (Lai & Sheu, 2017). In the
Italian state there is also a study of the polarization of
x-gamma rays produced by Thomond an Compton
scattering. (Petrillo et al., 2015). In Ridgers research, in
high-intensity laser interference (> 1021 Wcm-2) the
emission of gamma-ray photons by electrons can
greatly affect the dynamics of electrons and the
excessive number of electron-positron pairs can be
produced by emitted photons (Ridgers et al., 2014) .
In the BNCT experiment requires a neutron
source. This is because neutron sources have
important factors that produce flux and energy.
Generally, this method is used in reactors that have a
neutron source for treatment (Heydari & Ahmadi,
2015). In Indonesia there are three nuclear research
reactors operated by Atom and Nuclear Agencies
National (BATAN). The reactors are TRIGA 2000
reactor in Bandung, GA Siwabessy Multipurpose in
Serpong, and TRIGA MARK II (Kartini Reactor) in
Yogyakarta (Priambodo, Nugroho, Palupi, Zailani, &
Sardjono, 2017). Kartini reactor is one of TRIGA
reactor (Training Isotope Production by General
Atomic). The reactor as a nuclear reactor is still
operating with 100kW power. The reactor has many
experimental facilities located in Lazy Susan,
Pneumatic Transfer system, 2 radial beamports,
tangetial beamport, thermal column, and radial
piercing beamport. Beamport includes one facility of
a reactor that has a neutron flux higher than the other.
Then in front of the radial piercing beamport hole
there must be a shielding that serves to absorb the
neutron and gamma radiation coming out of the
neutron source. (Widarto, 2016).
The radiation shield is a combination of
radioactive sources aimed at reducing radiation. The
materials used are materials that have density and
homogeneity composition(Lakshminarayana et al.,
2017). The most important factor for reducing this
effect, is determining the most adequate material for
the shield (Elmahroug, Tellili, & Souga, 2014). Based
on the analysis and theory, the material used to
protect high energy neutrons is 14 MeV heavy metals
that contain lots of hydrogen. In elastic scattering
materials commonly used are polyethylene, paraffin,
water and polyethylene boron (Ding et al.,
2015).Thermal neutrons produced with low energy,
have a much greater absorption rate (Malkapur et al.,
2017).
Based on James Chadick's experiment on
pene, the manganese neutron paraffin is the most
widely used material on nuclear facilities (Toyen &
Saenboonruang, 2017). The optimum amount of
paraffin can accelerate rapid neutrons. And to
minimize the dispersion of the wall, the shield is laid
4He + 7 Li ( 2.79
MeV) 6.1%
4He + 7 Li ( 2.31 MeV)
+
γ (0.48 MeV ) 94%
Fig 1. The principles of working of BNCT
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33
down a mere 4m from the neutron source (Waheed et
al., 2017). Whereas Pb has a higher number of atoms
in the periodic table so it is good to protect gamma
radiation and there is another addition of aluminum
oxide used to increase mechanical strength (Kaur &
Singh, 2014) Based on studies of the interaction
between anat neutrons and materials, shielding
materials used must contain low atomic elements
such as C and H, high neutron cross sections such as
B and Gd, and high atomic numbers such as Pb and
W(Zhang et al., 2017). Neutron attenuation is
performed through elastic and inelastic scattering
reactions that aim to reduce neutron energy until it is
absorbed. The neutron catch section is larger for
thermal neutron energy only. Therefore, neutrons that
slow down by scattering are essential before being
caught (Jasim & Abdulameer, 2014). The neutron
damping parameters are the neutron reduction factors
between the first foil location and the foil location
respectively (Nyarku, Keshavamurthy, Subramanian,
Haridas, & Glover, 2013)
In document no. In 1990, the International
Commission for Radiological Protection (ICRP), a
comprehensive dose-limiting system should be
adopted. Radioactive substances and other radiation
sources are based on the principle of benefit and must
first be approved by the Supervisory Board (the
principle of justification), all radiation should be kept
as low as possible (Sari, Sardjono & Widiharto,
2017).According to the regulation of the Head of
Nuclear Power Supervisory Agency No. 4 of 2013 in
chapter III chapter 15 there are several application of
radiation protection requirements, namely:
a) Effective doses averaging 20 mSv (twenty
milisievert) per year within a period of 5
(five) years, so that Doses accumulated in 5
(five) years should not exceed 100 mSv (one
hundred milisievert);
b) Effective dose of 50 mSv (fifty milisievert) in
1 (one) year;
c) The equivalent dose for an average eyepiece
of 20 mSv (twenty milisievert) per year
within a period of 5 (five) years and 50 mSv
(fify milisievert) in 1 (one) year;
d) Equivalent dose for the skin of 500 mSv (five
hundred milisievert) per year; and
e) The equivalent dose for the hands or feet of
500 mSv (five hundred milisievert) per year.
(BAPETEN, 2013)
Monte Carlo simulation is a method to
simulate the statistical system. (Walter & Barkema,
2015). In this study, Monte Carlo calculations were
performed for gamma shield design. Purpose of
design:
1. Maximize neutron moderation to increase neutron
thermal flux which results in increased prompts of
gamma ray flux of sample material,
2. Minimize rapid and epithermal neutron flux,
3. Minimize gamma rays emitted from the source and
any delayed reaction (Hadad, Nematollahi,
Sadeghpour, & Faghihi, 2016).
MCNP is a general purpose, continuous
energy, general geometry, time dependent combined
with monte calo transport code. Where modes of
transport that can be used are neutrn, photon,
electron, displacement of neutron / photon. Nutron
energy is 10-11 MeV, 20 for all isotopes and 150
MeV for some isotopes. And the photon energy is 1
keV to 100 GeV(X-5 Monte Carlo Team,2005). The
Monte Carlo model has been developed using
MCNP5 to simulate the activation process (Ródenas,
2017). However there is a new version of MCNP that
is MCNP Extended (MCNPX) (Sardjono, 2015). The
MCNPX simulation results show that the epithermal
neutron flux released by the collimator meets the
IAEA standard is equal to 1.02241x1010 n / cm2 -
s(Yuniarti & Sardjono, 2016).
2. EXPERIMENT AND METHOD
This study aims to model shielding and
determine the rate of radiation exposure dose in the
BNCT facility area in vitro in vivo test. The material
used is Paraffin and Lead (Pb). The instrument used
in this research is portable computer hardware. The
software used is notepad ++, MCNPX, command
prompt, and visual editor (vised). Research begins by
collecting data on BNCT facilities.
The maximum effective dose limit required
by BAPETEN is 20 mSv per year. The calculation
assumption used is with aspect the most conservative,
that is the length of the worker in one year and 1920
hoursis positioned right on the surface of shielding.
Calculation of the dose rate is :
Ḣ = 20000 µSv/year
Ḣ = 20000 µSv/year : 1920 hours/year
Ḣ = 10.42 µSv/hours
There are variables used in this study
including independent variables, bound variables, and
controlled variables. The procedure in this study is
literature study, shielding modeling using mcnpx,
determining the flux coming from the mouth of the
collimator, obtaining the flux through the paraffin
shielding, and the lead casing. After obtaining the
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e-ISSN 2550-0570, © FSM UKSW Publication
34
obtained flux it is converted to dose rate out of BNCT
facility.
In shielding modeling and dose rate
determination using MCNPX. First create a geometry
by inputting the input on notepad. Determination of
flux obtained using tally F4. Tally is used to record
the neutron energy coming out from the reactor core
on the thermal column. The code of the input is as
follows:
1. Cell Card
Tabel 1. Input cell card
Parameter
input Description
J Cell number; 1 ≤ 𝑗 ≤ 99999.
M Material number
D
The density number of the
material used. Negative value
for mass density and positive
value for atomic density
Geom specifications of geometry
Params Optional specifications of cell
parameters
2. Surface Card
The boundaries of that geometry
is an expression of mathematical equations.
3. Data Card
Some data cards require a pointer to distinguish
data inputs for neutrons, photons, and electrons. Data
cards are divided into the following categories:
MP:N : cell dan parameter surface
SDEF : source specification
Fn, En : tally specification
Mn : Material specification
NPS : Problem cutoffs
(Briesmeister, 2000).
The code is used when determining the
current flow rate, flux, and dose rate. Measurement of
this dose rate is obtained from running MCNPX with
tally f4. Codes that have been created in the notpad
are saved with * .i format. Then the program is run
using command prompt to obtain data rate dose using
tally F4. The researchers tested the material to
determine the ability of the material by running a
shielded source with a certain thickness. Gy / s dose
rate is obtained from MCNPX output. Then converted
to µSv/h by multiplying each particle by weight
factor.
Table 2. Radiation weight factor
Radiation Weight factor
Photon 1
Neutron
E<500 keV 5
0.5 MeV<E<1 MeV 10
1 MeV<E<2.5 MeV 13
2.5 MeV<E<20 MeV 20
(Tsurayya,2017)
3. RESULT AND DISCUSSION
Based on the analysis that has been done,
then obtained modeling shielding made from paraffin
as follows:
Fig 2. Shielding Paraffin Design
In the shielding design there are some
materials used are Paraffin shown in red and lead are
shown in blue. The neutron source is indicated by the
yellow color and the green color indicates soft tissue.
Shielding that has been made has a thickness of 40
cm, the casing made of Pb has a thickness of 25cm
and the outer soft tissue has a thickness of 5 cm. This
neutron source has an energy of 14 MeV then goes
into the irradiation chamber and the neutron interacts
with the material to produce a flux of 333.31 μSv / h.
Neutrons in addition to scattering can also
experience an absorption reaction. The reaction can
produce gamma radiation (Lamarsh, 1961). This
gamma radiation is able to penetrate paraffin,
therefore added Pb material to absorb the gamma
radiation. Because Pb has a higher number of atoms
Indonesian Journal of Physics and Nuclear Applications, Vol.5, No.2, June 2020
35
in the periodic table so it is good to protect the
gamma radiation ((Kaur & Singh, 2014)). The
resulting radiation after passing Pb becomes 6.5 μSv /
h.
Fig 3. Dose Rate Radiation
Fig 3. above is a graph of dose rate on paraffin
shielding with a thickness of 40 cm. On the graph can
be seen the dose reduction outside BNCT facilities.
At 41.28 μSv / h thickness Pb 5cm, 19.12 μSv/h
thickness Pb 10cm, and dose rate with value 6.5
μSv/h has a thickness of 25 cm. On the assumptions
used in radiation protection ie the working time of
radiation in one year is 1920 hours with the exact
distance beyond the surface of shielding. The
maximum dose limit that workers receive is 10.42
μSv / hr. Meanwhile, based on the result of program
simulation is 6.5 μSv/h.
4. CONCLUSION
According to the regulation of the Head of
Nuclear Power Supervisory Agency No. 4 of 2013 in
chapter III chapter 15 there are several applications of
radiation protection requirements, effective doses
averaging 20 mSv (twenty milisievert) per year
within a period of 5 (five) years. While the maximum
dose limit received by workers is 10.42 μSv / h. In
shielding simulation result using MCNPX software
resulted dose rate of radiation exposure outside
BNCT facility in vitro in vivo test that is equal to 6.5
μSv / h. The thickness of shielding paraffin used is 40
cm, PB casing 25 cm, and 5cm soft tissue.
ACKNOWLEDGE
We would like to thank all members who supporterd
this project. We would especially like to thank the
Center for Accelerator Science and Technology
(PSTA BATAN) for the opportunities given to
perform this work.
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