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Indonesian Journal of Physics and Nuclear Applications

ijpna
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Published by Universitas Kristen Satya Wacana

ISSN : 2549046X EISSN : - DOI : -

Core Subject : Science, Social,

Environmental Science Physics

Indonesian Journal of Physics and Nuclear Applications is an international research journal, which publishes top level work from all areas of physics and nuclear applications including health, industry, energy, agriculture, etc. It is inisiated by results on research and development of Indonesian Boron Neutron Capture Cancer Therapy (BNCT) Consortium. Researchers and scientists are encouraged to contribute article based on recent research. It aims to preservation of nuclear knowledge; provide a learned reference in the field; and establish channel of communication among academic and research expert, policy makers and executive in industry, commerce and investment institution.

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Articles 80 Documents

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

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 Sri Yuniarti; Yohannes Sardjono; Bilalodin Bilalodin
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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) Aniti Payudan; Abdullah Nur Aziz; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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 BNCT'S 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 IAEA'S 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 thermal's 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 Rosenti Pasaribu; Kusminarto Kusminarto; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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 Martinus I Made Adrian Dwiputra; Andang Widi Harto; Yohannes Sardjono; Gede Sutisna Wijaya
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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 Ranti Warfi; Andang Widi Harto; Yohannes Sardjono; Widarto Widarto
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 1 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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 Adrian Tesalonika; Andang Widi Harto; Yohannes Sardjono; Isman Mulyadi Triatmoko
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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 Anggraeni Dwi Susilowati; Kusminarto Kusminarto; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

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.

The Application of Nuclear Medicine Giner Maslebu; Suryasatriya Trihandaru
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

Currently, the practice of nuclear medicine in modern countries comprises a large number of procedures. It is applied to study function of organs/body systems, to visualize, to characterize, and to quantify the functional state of lesions and for targeted radionuclide therapy. This overview presents all kinds of application in nuclear medicine services. Instrumentation and radioactive/radiolabeled substances are the basic components for application. Biotechnology contributes to the development and production of biomolecules used in radiopharmaceuticals. As a diagnostic modality, imaging depicts radioactivity distribution as a function of time. Hybrid imaging provides more precise localization and definition of le-sions as well as molecular imaging cross validation. Counting tests study invivoorgan functions externally or assess analytes in the biologic samples. Radiopharmaceutical therapy can be applied directly into the lesion or targeted systemically. Nanotechnology facilitates targeting and opens the development of bimodal techniques. In addition, neutron application contributes to the advancement of nuclear medicine services, such as neutron activation analysis, neutron teletherapy and neutron capture therapy.

Identification of Heavy Metal in Palm Oil Empty Fruit Bunch Compost, Mulch from Palm oil Waste and Its Effect on Chili (Capsicum annuum L.) Mercy Bientri Yunindanova; Herdhata Agusta; Dwi Asmono
Indonesian Journal of Physics and Nuclear Applications Vol 1 No 2 (2016)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

This study aimed to investigate the effect of compost from oil palm empty fruit bunches with different ages, mulching oil palm waste, the levels of heavy metals in compost and its influence on the growth and yield of chili. Composting was assisted by the starter (PromiTM) with a dose of 0.5 kg per ton of chopped bunches. Composting treatment distinguished by the composting time namely 4, 6, 8 and 10 weeks. Mulch treatment consisted of shell, fiber and empty fruit bunch chopped. The empty fruit bunch compost had the potential to decrease the soil acidity because the pH of 7.89- 8.66. The EFBs compost contained Boron of 7.7-10.7 ppm, 12-24.8 ppm of Cuprum, 0.05 to 0.24 % of Fe, 26.5-89.7 ppm of Mn, and 9.1-10.8 ppm of Na. This compost contained heavy metal Cd and Hg. Cd was detected in amount of 0.08 to 0.25 ppm. Hg was detected in amount of 12.9-19.5 ppm. Meanwhile, Pb and As were not found. Cd did not exceed the threshold. On the other hand, Hg was detected exceeding the threshold but did not affect the growth and yield of chili. Organic mulch from palm oil wastes did not significantly affect on the chili yield. Shell mulch had a negative influence on the growth and yield of chili.

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