modelling of paraffin shielding for bnct facility at

31

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

mailto:[email protected]

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

Indonesian Journal of Physics and Nuclear Applications, Vol.5, No.2, June 2020

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



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


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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