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Journal : Indonesian Journal of Physics and Nuclear Applications

A Design of Boron Neutron Capture Therapy for Cancer Treatment in Indonesia Sardjono, Yohannes; Widodo, Susilo; Irhas, Irhas; Tantawy, Hilmi
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

Boron Neutron Capture Therapy (BNCT) is an advanced form of radiotherapy technique that is potentially superior to all conventional techniques for cancer treatment, as it is targeted at killing individual cancerous cells with minimal damage to surrounding healthy cells. After decades of development, BNCT has reached clinical-trial stages in several countries, mainly for treating challenging cancers such as malignant brain tumors. The Indonesian consortium of BNCT already developed of the design BNCT for many cases of type cancers using many neutron sources. The main objective of the Indonesian consortium BNCT are the development of BNCT technology package which consists of a non nuclear reactor neutron source based on cyclotron and compact neutron generator technique, advanced boron-carrying pharmaceutical, and user-friendly treatment platform with automatic operation and feedback system as well as commercialization of the BNCT though franchised network of BNCT clinics worldwide. The Indonesian consortium BNCT will offering to participate in Boron carrier pharmaceuticals development and testing, development of cyclotron and compact neutron generators and provision of neutrons from the 100 kW Kartini Research Reactor to guide and to validate compact neutron generator development. Studies were carried out to design a collimator which results in epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle 5 (MCNP5) codes. Reactor within 100 kW of output thermal power was used as the neutron source. The design criteria were based on the IAEA’s recommendation. All materials used were varied in size, according to the value of mean free path for each. Monte Carlo simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 1,5 cm thick of Bi as "-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 3-5 cm varied aperture size, epithermal neutron beam with minimum flux of 7,8 x 108 n.cm-2.s-1, maximum fast neutron and "-ray components of, respectively, 1,9 x 10-13 Gy.cm2.n-1 and 1,8 x 10-13 Gy.cm2.n-1, maximum thermal neutron per epithermal neutron ratio of 0,009, and beam minimum directionality of 0,72, could be produced. The beam did not fully pass the IAEA’s criteria, since the epithermal neutron flux was still below the recommended value, 1,0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeded 5 x 108 n.cm-2.s-1. When this collimator was surrounded by 8 cm thick of graphite, the characteristics of the beam became better that it passed all IAEA’s criteria with epithermal neutron flux up to 1,7 x 109 n.cm-2.s-1. it is still feasible for BNCT in vivo experiment and study of many cases cancer type i.e.; liver and lung curcinoma. In this case, thermal neutron produced by model of Collimated Thermal Column Kartini Research Nuclear Reactor, Yogyakarta. Sodium boroncaptate (BSH) was used as in this research. BSH had effected in liver for radiation quality factor as 0.8 in health tissue and 2.5 in cancer tissue. Modelling organ and source used liver organ who contain of cancer tissue and research reactor. Variation of boron concentration was 20, 25, 30, 35, 40, 45, and 47 $g/g cancer. Output of MCNP calculation were neutron scattering dose, gamma ray dose and neutron flux from reactor. Given the advantages of low density owned by lungs, hence BNCT is a solid option that can be utilized to eradicate the cell cancer in lungs. Modelling organ and neutron source for lung carcinoma was used Compact Neutron Generator (CNG) by deuterium-tritium which was used is boronophenylalanine (BPA). The concentration of boron-10 compound was varied in the study; i.e. the variations were 20; 25; 30; 35; 40 and 45 μg.g-1 cancer tissues. Ideally, the primary dose which is solemnly expected to contribute in the therapy is alpha dose, but the secondary dose; i.e. neutron scattering dose, proton dose and gamma dose that are caused due to the interaction of thermal neutron with the spectra of tissue can not be simply omitted. Thus, the desired output of MCNPX; i.e. tally, were thermal and epithermal neutron flux, neutron and photon dose. The liver study variation of boron concentration result dose rate to every variation were0,042; 0,050; 0,058; 0,067; 0,074; 0,082; 0,085 Gy/sec. Irradiation time who need to every concentration were 1194,687 sec (19 min 54 sec);999,645 sec (16 min 39 sec); 858,746 sec (14 min 19 sec); 743,810 sec (12 min 24 sec); 675,156 sec (11 min 15 sec); 608,480 sec (10 min 8 sec); 585,807sec (9 min 45 sec). The lung carcinoma study variations of boron-10 concentration in tissue resulted in the dose rate of each variables respectively were 0.003145, 0.003657, 0.00359, 0.00385, 0.00438 and 0.00476 Gy.sec-1 . The irradiated time needed for therapy for each variables respectively were 375.34, 357.55, 287.58, 284.95, 237.84 and 219.84 minutes.
Design Collimator and Dosimetry of in Vitro and in Vivo Test Using MCNP-X Code Yuniarti, Sri; Sardjono, Yohannes; Bilalodin, Bilalodin
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

Studies were carried out to collimator modelling and dosimetry BNCT of in vitro and in vivo test using MCNP-X code. Collimator modelling performed to obtain neutron beam as required by the International Atomic Energy Agency (IAEA). Dosimetry calculations performed to obtain the results of the dose calculation (dosimetry) in the application of BNCT.  Collimator modelling and dosimetry simulations performed with MCNPX program. Neutron sources used for simulation, namely cyclotrons HM-30, energy 30 MeV, the current is 1.1 mA. Collimator modelling utilizes to program MCNPX covers cells target as beryllium, collimator wall (reflector), moderate, filter, gamma-ray shielding, and aperture. The simulation results of the modelling are Φepi 1.02241x1010 n/cm2 s, Df/Φepi 2.36487x10-11 Gy-cm2/n, Dγ/Φepi 4.68416x10-12 Gy-cm2/n, Φth/Φepi 3.76285x10-01, J/Φepi 8.37678x103. Based on the calculation of the dose rate that has been done, the result that the optimal dose rate at a depth of 1cm.
Basic Principle Application and Technology of Boron Neutron Capture Cancer Therapy (BNCT) Utilizing Monte Carlo N Particle 5’S Software (MCNP 5) with Compact Neutron Generator (CNG) Payudan, Aniti; Aziz, Abdullah Nur; Sardjono, Yohannes
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

The purpose are to know basic principle, needed component, types of compact neutron generator, plus and minus CNG, identify materials can use as collimator, know physics parameters as input software MCNP 5, knowing step simulation with software MCNP 5, dose in BNCT, knowing boron compound that use in BNCT, getting collimator design for BNCTS application with source is compact neutron generator and count physics parameter of collimator output and compares it with standard IAEA. Method are reading reference and simulation with MCNP 5. The result are BNCT use high linear energy transfer from alpha and lithium as a result of 10B(n,α)7Li reaction. BNCT method is effective for cancer therapy. It is not dangerous to normal tissues. To work perfectly, BNCT needs neutron, boron (BSH and BPA as boron compound) Indonesia have study turmeric as boron compound, neutron source, collimator and dose. Dose component in BNCT that important are dose of recoil proton, dose of gamma, dose alfa and dose radiation to environmentally. CNG produce neutron with fussion reaction of deuterium-deuterium (2,45 MeV), deuterium-tritium (14 MeV), tritium-tritium(11,31 MeV) can used as neutron source BNCT. Many kinds of CNG are axial, coaxial, toroidal, plasma design, accelerator design, and CNG with diameter 2,5 cm. CNG have more benefit than another neutron source, make CNG compatible as BNCT application. Neutron from CNG need collimator to get neutron as IAEA’s parameter.  Material for collimator are wall and aperture (material: Ni, Pb, Bi), moderator (Al, Al2O3, S, AlF3), filter (6Li,10B, LiF, Al, Cd-nat,  Ni-60, BiF3, 157Gd, 151Eu), gamma shield (Bi, Pb). Simulation using MCNP 5 has severally steps, the first is sketching problem, the second is making listing program with notepad, the third open program on visual editor, and the last is running program. Acquired result is design tube collimator with radius 71 cm and high 139, 5 cm. Design contained on lead wall as thick as 19, 5 cm; moderate: heavy water as thick as 4 cm, AlF3 girdle a half of part CNG, MgF 2 (19 cm + 10 cm), Al (6,5 cm + 5 cm);Gamma shield: bismuth, and aperture with diameter 6 cm by steps aside nickel. The result collimator output cross three of five IAEAS defaults. They are the ratio among dosed gamma with flux epithermal is 5,738×10 -24Gy. cm 2 .n -1, the value of ratio among thermals neutron flux with epithermal neutron is 0, 02567, and ratio among current with flux neutron completely is 1, 2. Need considerable effort of all part to realize BNCT in Indonesia.
Clinical trial design of Boron Neutron Capture Therapy on breast cancer using D-D coaxial compact neutron generator as neutron source and Monte Carlo N-Particle simulation method Pasaribu, Rosenti; Kusminarto, Kusminarto; Sardjono, Yohannes
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

A clinical trial simulation of Boron Neutron Capture Therapy (BNCT) for breast cancer was conducted at National Nuclear Energy Agency Yogyakarta, Indonesia. This was motivated by high rate of breast cancer in the world, especially in Indonesia. BNCT is a type of therapy by nuclear reaction 10B(n,α)7Li that produces kinetic energy totaling 2.79 MeV. High Linear Energy Transfer (LET) radiation of α-particle and recoil 7Li would locally deposit their energy in a range of 5-9 μm, which corresponds to the human cell diameter. Fast neutron coming out of Compact Neutron Generator (CNG) was moderated using Fe and MgF2 material. A collimator, along with breast cancer and the corresponding organ at risk were designed compatible to Monte Carlo N-Particle X (MCNPX). The radiation were simulated by the MCNPX software and the physical quantities were counted by tally MCNPX codes. The highest neutron thermal flux was found at a depth of 1.4 cm on fat tissue. En face and upward intersection radiation techniques were adopted for the breast cancer radiation. The average dose rate of radiation used on breast cancer was 1.72×10-5 Gy/s for the en face method and 8.98×10-6 Gy/s for the upward intersection method. Dose 50±3 Gy was given into cancer cell, (4.18±0.06) ×10-2 Gy into heart and (8.16±0.06) ×10-2Gy into lung for 806.34 hours irradiation.
Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator Dwiputra, Martinus I Made Adrian; Harto, Andang Widi; Sardjono, Yohannes; Wijaya, Gede Sutisna
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

Studies were carried out to design a shielding for BNCT facility in the end of Kartini reactor’s thermal column with predesigned collimator. The design consist of selecting the material and their thickness. The shielding is required to absorb the leaking radiation until the Dose Limit Value of 20 mSv/year for radiation worker is met. The material considered were paraffin, barite concrete, borated polyethylene, stainless steel 304 and lead. The calculation was done using MCNPX tally facility with converted dose limit value of 10.42 µSv/hour. Design number two were chosen as the best from three designs which surrounded a room with length, width and height of, respectively 200 cm, 200 cm and 166.4 cm. The first and main layer are borated polyethyelene and barite concrete of 20 and 30 cm, respectively. The additional layer are borated polyethyelene and barite concrete of 15 cm and 15 cm with less volume than the main layer to decrease the primary straight radiation from the thermal column. Maximum radiation dose rate is 7.0746 µ Sv/hour in cell 227 with average dose rate of 2.58712 µSv/hour.
Optimization of Neutron Collimator in The Thermal Column of Kartini Research Reactor for in vitro and in vivo Trials Facility of Boron Neutron Capture Therapy using MCNP-X Simulator Warfi, Ranti; Harto, Andang Widi; Sardjono, Yohannes; Widarto, Widarto
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

The optimization of thermal column collimator has been studied which resulted epithermal neutron beam for in vivo and in vitro trials of Boron Neutron Capture Therapy (BNCT) at Kartini Research Reactor of 100 kW by means of Monte Carlo N-Particle Extended (MCNP-X) codes. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA). MCNP-X calculations indicated by using 5 cm thickness of Ni as collimator wall, 30 cm thickness of Al as moderator, 20 cm thickness of 60Ni as filter, 2 cm thickness of Bi as γ-ray shielding, 3 cm thickness of 6Li2CO3-polyethylene as beam delimiter, and for in vivo in vitro trials purpose, aperture was designed 8 cm radius size, an epitermal neutron beam with an intensity 1.13E+09 n.cm-2.s-1, fast neutron and γ-doses per epithermal neutron of 1.76E-13 Gy.cm2.n-1 and 1.45E-13Gy.cm2.n-1,minimum thermal neutron per epithermal neutron ratio of 0.008,and maximum directionality of 0.73, respectively could be produced. The results have passed all the IAEA’s criteria.
Dosimetry of in vitro and in vivo Trials in Thermal Column Kartini Reactor for Boron Neutron Capture Therapy (BNCT) facility by using MCNPX Simulator Code Tesalonika, Adrian; Harto, Andang Widi; Sardjono, Yohannes; Triatmoko, Isman Mulyadi
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

A dosimetry study of in vitro and in vivo trials system in thermal column of Kartini Reactor for Boron Neutron Capture Therapy (BNCT) facility has been conducted by using the Monte Carlo N-Particle Extended (MCNPX) software. Source of neutron originated from the 100 kW reactor which has been modified by the previous researcher. Models have been made by using simple geometries to represent tissues. Models of in vitro have been made by 4 spheres each has 1 cm diameter to represent tumour, whereas in vivo by 4 cylinders each has 6 cm length, 3 cm diameter, and breast soft tissue material with 1 cm sphere each located in the center of the cylinders to represent models of mouse with tumour. An increase in value of the boron concentration will increase the value of dose rate as well, then the exposure time should be shorter. The exposure times (in minutes) of in vitro trials for 20, 25, 30, 50, 75, 100, 125, and 150 μg boron/g tissues are 117.77, 117.77, 117.07, 115.69, 114.02, 112.39, 110.80, and 109.27. Whereas the exposure times of in vivo trials are 163.58, 162.78, 161.98, 158.88, 155.16, 151.61, 148.22, dan 144.98. In vitro trials have greater values of dose rate so that in vitro trials have shorter exposure time.
Boron Neutron Capture Therapy (BNCT) using Compact Neutron generator Susilowati, Anggraeni Dwi; Kusminarto, Kusminarto; Sardjono, Yohannes
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

Boron Neutron Capture Therapy (BNCT) must be appropriate with five criteria from IAEA. These criteria in order to prevent neutron beam output harm the patient. It can be by using Collimator of neutron source Compact Neutron Generator (CNG) and Monte Carlo simulation method with N particles 5 .CNG is developed by deuteriumtritium reaction (DT) and deuterium-deuterium (DD) reaction. The manufacture result of the collimator is obtained epithermal neutron flux value of 1.69e-9 n/cm^2s  for D-T reaction and 8e6 n/cm^2s for D-D reaction, ratio of epithermal and thermal is 1.95e-13 Gy cm^2/n for D-T reaction and for D-D reaction, ratio of fast neutron component is 1.69e-13 Gy cm^2/n for D-T reaction and for D-D reaction, ratio of gamma component is 1.18e-13 Gy cm^2/nfor D-T reaction and for D-D reaction. The Latest reaction is current ratio 0.649 for D-T reaction and 0.46 for D-D reaction.
Preparation of Dosimetry of Boron Neutron Capture Therapy (BNCT) for In vivo Test Planning system using Monte Carlo N-Particle Extended (MCNP-X) Software Arrozaqi, Muhammad Ilma Muslih; Kusminarto, Kusminarto; Sardjono, Yohannes
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

Cancer is a disease with second largest patients in the world.  In Indonesia, the number of radiotherapy facility in Indonesia is less than 30 units and every patients needs more than single exposure, so that it result a long waiting list of treatment up to one year. Now, a new treatment of cancer is developed. It is Boron Neutron Capture Therapy that using capture reaction of neutron by Boron-10. Before this method is applied to patient, it requires some testing which is one of them is in vivo test. This research has been conducting to prepare the in vivo test, especially in dosimetry. Preparation of dosimetry includes collimator design and mouse phantom model. The optimum specification of the collimator is consist of Nickel collimator wall with 2 cm of thickness, Aluminum moderator with 10 cm of thickness and lead gamma shield with 3.5 of thickness. This design result in 1.18 x 108 n/cm2s of epithermal and thermal neutron flux, 2,24 x 10-11 Gy cm2/s of fast neutron component dose, 1.35 x 10-12 Gy cm2/s of gamma dose component, and 7.18 x 10-1 of neutron current and flux ratio. Mouse phantom model is built by two basic kind of geometry, they are Ellipsoid and Elliptical Tory. Both of basic geometry can be used to make all important organs of mouse phantom for dosimetry purpose.
Optimization of a Beam Shaping Assembly Design for Boron Neutron Capture Cancer Therapy Facility Based on 30 MeV Cyclotron Ardana, I Made; Kusminarto, Kusminarto; Sardjono, Yohannes
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 3 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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Abstract

A series of simulations has been carried out using a Monte Carlo N Particle X code to find out the final composition and configuration of a neutron Beam Shaping Assembly (BSA)  to moderate the fast neutron flux which is generated from the thick disk-type beryllium target. The final configuration for neutron BSA design included 35 cm lead as reflector, 39 cm alumina as moderator, 8.2 cm lithium fluoride as fast neutron filter and 0.5 cm boron carbide as thermal neutron filter. Bismuth, lead fluoride, and lead were chosen as the aperture, reflector, and gamma shielding, respectively. The disk-type of beryllium target is 19 cm in diameter with 0.5 cm thickness which is covered by copper plate to hold the water pressured coolant. A higher yield of neutron production requires a higher intensity of proton beams, which generate much heats and causes the target material to melt. Therefore, it is useful to consider the temperature distribution on the target material with flowing water coolant by means of computer modeling while designing the target. ANSYS-Fluent code will be used to estimate the thermal transfer and heat calculation in a solid target during beam irradiation. Epithermal neutron flux in the suggested design were 1,03x109 n/cm2 s, with almost all IAEA parameters for BNCT BSA design has been satisfied.
Co-Authors Abdullah Nur Aziz Adrian Tesalonika, Adrian Agung Prastowo, Agung Andang Widi Harto Andang Widiharto Anggraeni Dwi Susilowati, Anggraeni Dwi Aniti Payudan, Aniti Arief Hermanto Aulia Setyo Wicaksono, Aulia Setyo Bemby Yulio Vallenry Bilalodin Bilalodin Bima Caraka Putra, Bima Boni Pahlanop Lapanporo Budi Setyahandana Darmayanti, Alifia Dwi Satya Palupi Eko Priyono Fahrudin Nugroho Fajar Nurjaman Faqqiyyah, Hamidatul Fasni, Bagus Novrianto Ferdy S. Rondonuwu Gede Bayu Suparta Gede Sutisna Wijaya, Gede Sutisna Giner Maslebu, Giner Harish, Ahmad Faisal Hasyim, Kholidah Hilmi Tantawy, Hilmi I Made Ardana Irhas Irhas, Irhas Isa Akhlis Isman Mulyadi Triatmoko, Isman Mulyadi Jans P B Siburian, Jans P B Jodelin Muninggar, Jodelin Kusminarto Kusminarto Larry E. Fennern Larry E. Fennern M. Ibnu Khaldun, M. Ibnu Mahmudah, Rida Siti Nur’aini Martinus I Made Adrian Dwiputra, Martinus I Made Adrian Masanori Aritomi Maysaroh, Atika Mu’Alim, Muhammad Muhammad Ilma Muslih Arrozaqi, Muhammad Ilma Muslih Nina Fauziah Ntoy, Suhendra Gunawan Nur Endah Sari, Nur Endah Prayoga Isyan, Prayoga Priambodo, Gani Ralind Re Marla Ranti Warfi, Ranti Ratnasari, Nunung Gupita Rosenti Pasaribu, Rosenti S.B., Ramadhan Valiant Gill Sigit Santosa Siti Rosidah Slamet Parmanto Soeparmi Soeparmi, Soeparmi Sri Yuniarti Sunardi Sunardi Sunarno Sunarno Sungkowo Wahyu Santoso Supardi Supardi Suryasatriya Trihandaru Susilo Susilo Susilo Widodo, Susilo Syamputra, Dhani Nur Indra Syarip Syarip Wahyuni, Nur Setyo Warsono Warsono Widarto Widarto Wusko, Ikna Urwatul Yosaphat Sumardi Yuliana Dian N, Yuliana Dian Zailani, Rosilatul