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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 5 Documents

Search results for , issue "Vol 2 No 2 (2017)" : 5 Documents clear

An Optimization Design of Collimator in The Thermal Column of Kartini Reactor For BNCT M. Ibnu Khaldun; Andang Widi Harto; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 2 No 2 (2017)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

Studies were carried out to design a collimator which results in epithermal neutron beam for in vivo experiment of Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 6 cm thick of Natural Nickel as collimator wall, 65 cm thick of Al as moderator, 3 cm thick of Ni-60 as filter, 6 cm thick of Bi as γ-ray shielding, 3.5 cm thick of Li2CO3-polyethilene, with 2 cm aperture diameter. Epithermal neutron beam with maximum flux of 6.60 x 108n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.82 x 10-13Gy.cm2.n-1 and 1.70 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.041, and maximum directionality of 2,12. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. it is still feasible for BNCT in vivo experiment.

Calculation of Absorbed Dose Distribution for Breast Brachytherapy Simulation By CS-1 131Cs Seed and ADVANTAGETM 103Pd Seed Using Monte Carlo N Particle Extended Simulator Faisal Reza Rahmat; Mondjo Mondjo; Alexander Agung
Indonesian Journal of Physics and Nuclear Applications Vol 2 No 2 (2017)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

Simulation using Monte Carlo code has been conducted to determine the distribution of absorbed dose to the breast brachytherapy with 131Cs and 103Pd radionuclide sources. Simulations performed on stage I breast cancer with cancer diameter is 2 cm. Sources of radionuclides simulated in the form of seed is modeled with CS-1 which is made by IsoRay 131Cs and seed ADVANTAGETM103Pd which is made by IsoAID, LLC. Seed was planted in breast cancer cells. Calculation of absorbed dose distribution was performed by varying the distance from the seed. Variations of the distance started from a radius of 0.3 cm to 2 cm with a range of 0.1 cm respectively. In this simulation will also be reviewed the value of absorbed dose for healthy cell like breast, sternum, and lung. The relation between the absorbed dose and the distance from the seed can be described in the form of power law. The results of the calculation show that the maximum absorbed dose is in the target site of the cancer cells (5.791 ± 0.002) Gy per 5 MBq of 131Cs and (2.755 ± 0.009) Gy per 5 MBq for 103Pd. The absorbed dose at sternum (1.514 ± 0.011) x 10-4 Gy per 5 MBq of 131Cs and (7.515 ± 0.633) x 10-7 Gy per 5 MBq for 103Pd. While the absorbed dose in the lungs is and (3.615 ± 0.082) x 10-5 Gy per 5 MBq for 131Cs and (3.972 ± 0.591) x 10-8 Gy per 5 MBq for 103Pd.

Analysis of Radiation Effects on Workers and Environment Pilot Plant Boron Neutron Capture Therapy (BNCT) Nur Endah Sari; Yohannes Sardjono; Andang Widi Harto
Indonesian Journal of Physics and Nuclear Applications Vol 2 No 2 (2017)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

BNCT is a new method in nuclear technology. The aim of BNCT application is to reduce human risk which used to kills cell targeting characteristic. The impact of using this technology should be considered before it is applied, among the effects of radiation on workers and the surrounding environment BNCT pilot plant. A research on modeling of BNCT pilot plant used a collimator for a 30 MeV cyclotron neutron sources which had been designed from the past research. Radiation shielding modeling for treatment room used MCNPX software. The radiation shielding was concrete baryte on each side that includes coated borated polyethylene 2 cm thick and it is featured with a sliding door with dimensions 220 × 87 × 200 cm coated with stainless steels 2 cm thick. Results obtained value equivalent dose rate of neutron and gamma of each 41.5 µSv.h-1 and 2.05 µSv.h-1. Effects of radiation received by workers in the form of deterministic effects did not have a significant are impact.

A Conceptual Design Optimization of Collimator With 181Ta as Neutron Source for Boron Neutron Capture Therapy Based Cyclotron Using Computer Simulation Program Monte Carlo N Particle Extended Jans P B Siburian; Andang Widi Harto; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 2 No 2 (2017)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

The optimization of collimator with 30 MeV cyclotron as neutron source and 181Ta as its proton target. cyclotron assumed work at 30 MeV power with 1 mA and 30 kW operation condition. Criteria of design based on IAEA’s recommendation. Using MCNPX as simulator, the result indicated that with using 181Ta as target material with 0.55 cm thickness and 19 cm diameter, 25 cm and 45 cm PbF2 as reflector and back reflector, 30 cm 32S as a moderator, 20 cm 60Ni as fast neutron filter, 2 cm 209Bi as gamma filter, 1 cm 6Li2 CO3- polyethylenes as thermal neutron filter, and 23 cm diameter of aperture, an epithermal neutron beam with intensity 4.37 x 109 n.cm-2.s-1, fast neutron and gamma doses per epithermal neutron of 1.86 x 10-16 Gy.cm2.n-1 and 1.93 x 10-13Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.003, and maximum directionality 0,728, respectively could be produced. The results have passed all the IAEA’s criteria.

Internal Dose Analysis for Radiation Worker in Cancer Therapy Based on Boron Neutron Capture Therapy with Neutron Source Cyclotron 30 MeV Using Monte Carlo N Particle Extended Simulator Aulia Setyo Wicaksono; Andang Widi Harto; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 2 No 2 (2017)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

Based Studies were carried out to analyze internal dose for radiation worker at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and room that actually design before. This internal dose analyze include interaction between neutron and air. The air contains N2 (72%), O2 (20%), Ar (0.93%), CO2, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be bellow limit dose for radiation worker amount of 20 mSv/years. From the particle that are present in the air, only Nitrogen and Argon can change into radioactive element. Nitrogen-14 activated to Carbon-14, Nitrogen-15 activated to Nitrogen-16, and Argon-40 activated to Argon-41. Calculation using tally facility in Monte Carlo N Particle Version Extended (MCNPX) program for calculated flux Neutron in the air 3,16x107 Neutron/cm2s. room design in cancer facility have a measurement of length 200 cm, width 200 cm and high 166,40 cm. flux neutron can be used to calculated the reaction rate which is 80,1x10-2 reaction/cm3s for carbon-14 and 8,75x10-5 reaction/cm3s. Internal dose exposed to the radiation worker is 9.08E-9 µSv.

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