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All Journal International Journal of Power Electronics and Drive Systems (IJPEDS) Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega Teknofisika Indonesian Journal of Physics and Nuclear Applications Jurnal Manajemen Teknologi dan Teknik Sipil (JURMATEKS) Computational and Experimental Research in Materials and Renewable Energy (CERiMRE) Mitra Pendidikan: Jurnal Pendidikan dan Pembelajaran Movement and Education Journal of Integrated System
Andang Widiharto
Department Of Nuclear Engineering And Engineering Physics, Faculty Of Engineering, Universitas Gadjah Mada, Jalan Grafika 2, Yogyakarta 55281, Indonesia
Civil Engineering, Building, Construction & Architecture Computer Science & IT Control & Systems Engineering Education Electrical & Electronics Engineering Energy Engineering Environmental Science Industrial & Manufacturing Engineering Materials Science & Nanotechnology Physics Social Sciences Transportation Other

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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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (751.797 KB) | DOI: 10.24246/ijpna.v2i2.54-64

Abstract

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.
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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (395.904 KB) | DOI: 10.24246/ijpna.v2i2.75-82

Abstract

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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (949.26 KB) | DOI: 10.24246/ijpna.v2i2.83-90

Abstract

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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (720.824 KB) | DOI: 10.24246/ijpna.v2i2.91-100

Abstract

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.
Study on the Ability of PCMSR to Produce Valuable Isotopes as a By Product of Energy Generation Andang Widiharto
Indonesian Journal of Physics and Nuclear Applications Vol 3 No 1 (2018)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (268.282 KB) | DOI: 10.24246/ijpna.v3i1.7-14

Abstract

PCMSR (Passive Compact Molten Salt Reactor) is a variant of MSR (Molten Salt Reactor) type reactors. The MSR is one type of the Advanced Nuclear Reactor types. PCMSR uses mixtures of fluoride salt if the liquid form is in a high temperature operation. The use of liquid salt fuel allows the application of on line fuel processing system. The on line fuel processing system allows extraction of several valuable fission product isotopes such as Mo-99, Cs-137, Sr-89 etc. The capability of MSR to produce several valuable isotopes has been studied. This study is based on a denaturized breeder MSR design with 920 MWth of thermal power and 500 MWe of electrical output power with the thermal efficiency of 55 %. The initial composition of fuel salt is 70 % of a mole of LiF, 24 % of a mole of 232ThF4, 6 % of a mole of UF4. The enrichment level of U is 20 % of a mole of U-235. The study is performed by a numerical calculation to solve a set of differential equations of fission product balance. This calculation calculates fission product generation due to fission reaction, precursor decay, and fission product annihilation due to decay, neutron absorption, and extraction. The calculation result shows that in quasi equilibrium conditions, the reactor can produce several valuable isotopes in substantially sufficient quantities, those are Sr-89 (0.3 kCi/MWth/day, Sr-90 (1,91 Ci/MWth/day), Mo-99 (1.7 kCi/MWth/day), I-131 (0.42 kCi/MWth/day), I-132 (0.782 kCi/MWth/day), I-133 (1.12 kCi/MWth/day), Xe-133 (11.8 Ci/MWth/day), Cs-134 (39.3 mCi/MWth/day), Cs-137 (2.32 Ci/MWth/day) and La-140 (1.05 kCi/MWth/day).
DOSE OPTIMIZATION ON LIVER CANCER PROTON THERAPY AND BORON NEUTRON CAPTURE THERAPY USING PARTICLE AND HEAVY IONS TRANSPORT CODE SYSTEM Hafiz Fahrurrozi; Andang Widi Harto; Isman Mulyadi Triatmoko; Gede Sutrisna Wijaya; Yohannes Sardjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 1 (2021): FEBRUARY 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.1.6183

Abstract

Liver cancer was the third leading cause of death from cancer in 2020 with 830,180 deaths worldwide. Radiotherapy is a common treatment method for liver cancer. Technological advances presented proton therapy and boron neutron capture therapy (BNCT) as alternatives with a lower dose on healthy organs. The objective of this research is to get a good dose distribution with higher tumor dose and lower healthy organ dose in proton therapy. A comparison with BNCT is done to get a better understanding of how both methods deliver the dose to treat the cancer while minimizing healthy organ doses. The research simulated proton therapy for cancer liver with Particle and Heavy Ions Transport Code System (PHITS), and a literature review for BNCT. The effectiveness of both methods were compared by tumor dose and liver dose. The optimal tumor dose for proton therapy is 86.01 Gy (W) with 0.67 Gy (W) liver dose. Proton therapy can replace conventional radiotherapy for tumors with complex shapes in dose delivery by utilizing its dose profile, while BNCT can give better tumor control on patients previously treated with conventional radiotherapy.
ANALISIS TRANSIEN PADA FIXED BED NUCLEAR REACTOR M. Rizaal; Andang Widiharto; Sihana Sihana
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 13, No 3 (2011): Oktober 2011
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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Abstract

Desain teras Fixed Bed Nuclear Reactor (FBNR) yang modular memungkinkan pengendalian daya dapat dilakukan dengan mengatur ketinggian suspended core dan laju aliran massa pendingin. Tujuan penelitian ini adalah mempelajari perubahan daya termal teras sebagai akibat perubahan laju aliran massa pendingin yang masuk ke teras reaktor dan perubahan ketinggian suspended core serta mempelajari karakteristik keselamatan melekat yang dimiliki FBNR saat terjadi kegagalan pelepasan kalor (loss of heat sink). Keadaan neutronik teras dimodelkan pada kondisi tunak dengan menggunakan paket program Standard Reactor Analysis Code (SRAC) untuk memperoleh data fluks neutron, konstanta grup, fraksi neutron kasip, konstanta peluruhan prekursor neutron kasip, dan beberapa parameter teras penting lainnya. Selanjutnya data tersebut digunakan pada perhitungan transien sebagai syarat awal. Analisis transien dilakukan pada tiga kondisi, yaitu saat terjadi penurunan laju aliran massa pendingin, saat terjadi penurunan ketinggian suspended core, dan saat terjadi kegagalan sistem pelepasan kalor. Hasil yang diperoleh dari penelitian ini menunjukkan bahwa penurunan laju aliran massa pendingin sebesar 50%, dari kondisi normal, menyebabkan daya termal teras turun 28% dibanding daya sebelumnya. Penurunan ketinggian suspended core sebesar 30% dari ketinggian normal menyebabkan daya termal teras turun 17% dibanding daya sebelumnya. Sementara untuk kondisi kegagalan sistem pelepasan kalor, daya termal teras mengalami penurunan sebesar 76%. Dengan demikian, pengendalian daya pada FBNR dapat dilakukan dengan mengatur laju aliran massa pendingin dan ketinggian suspended core, serta keselamatan melekat yang handal pada kondisi kegagalan sistem pelepasan kalor.Kata kunci: FBNR, transien, daya, laju aliran massa, suspended core  Modular in design enables Fixed Bed Nuclear Reactor (FBNR) power controlled by the adjustment of suspended core and coolant flow rate. The main purposes of this paper are to learn the change of thermal power caused by the change of suspended core height and coolant flow rate, and also to learn the inherent safety when loss of heat sink condition prevailed. The Core was modelled on steady condition by using Standard Reactor Analysis Code (SRAC) to obtain neutron flux, group constants, delayed neutron fraction, delayed neutron precursor decay constants, and several core parameters. These data will be used as initial value on the transient calculations. Transient analysis was conducted on the following conditions: coolant flow rate changes, suspended core height changes and loss of heat sink occours. The calculated result showed that when the coolant flow rate is 50% decreased, thermal power of FBNR is 28% decreased. When suspended core height is 30% decreased, thermal power of FBNR is 17% decreased. Meanwhile, thermal power at loss of heat sink condition is 76% decreased. Therefore, the adjustment of suspended core height and coolant flow rate can control thermal power of FBNR, and FBNR’s inherent safety is reliable at loss of heat sink condition. Keywords: FBNR, transient, power, flow rate, suspended core
STUDI DESAIN DOWN SCALE TERAS REAKTOR DAN BAHAN BAKAR PLTN JENIS PEBBLE BED MODULAR REACTOR – HTR 100 MWe Slamet Parmanto; Andang Widiharto; Yohannes Sardjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 13, No 3 (2011): Oktober 2011
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (374.312 KB)

Abstract

Telah dilakukan penelitian terhadap teras reaktor Pebble Bed Modular Reactor (PBMR) dengan daya 100 Mwe berbahan bakar UO2. Reaktor ini menggunakan moderator grafit dan helium sebagai pendingin. Studi down scale dilakukan tanpa mengubah geometri teras maupun geometri bahan bakar. Parameter yang dianalisis adalah kritikalitas teras, reaktivitas lebih, koefisien reaktivitas temperatur bahan bakar, moderator dan pendingin serta nilai ekonomis bahan bakar. Dari penelitian ini diharapkan diperoleh desain bahan bakar yang bernilai ekonomis dan memiliki fitur keselamatan melekat. Penelitian dilakukan dengan menggunakan program SRAC 2003. Hasil yang diperoleh adalah desain bahan bakar UO2 berbentuk pebble dengan pengkayaan 10% U235 dan 90 ppm racun dapat bakar Gd2O3. Nilai faktor multipilkasi effektif keff pada beginning of life (BOL) adalah 1,01115 dan menjadi 1,00588 setelah 2658 hari operasi reaktor (EOL). Koefisien reaktivitas temperatur total diperoleh sebesar - 3,25900E-05 ∆k/k/K saat BOL dan -1,10615E-04 ∆k/k/K saat end of life (EOL). Reaktor ini memenuhi karakteristik keselamatan melekat ditandai dengan nilai koefisien reaktivitas temperatur yang negatif.Kata kunci: PBMR, desain bahan bakar, faktor multipilkasi effektif, reaktivitas lebih, koefisien reaktivitas temperatur. Research of Pebble Bed Modular Reactor (PBMR) 100 MWe which used UO2 fuel has been done. This reactor uses graphite as moderator and helium as coolant. Down scale studies performed without changing the core and fuel geometry. The parameter being analyzed were core criticality, excess reactivity, fuel, moderator, coolant temperature reactivity coefficient, and fuel economy. This research is expected to obtain the design that has fuel economy and inherent safety features. In this research, we have employed SRAC 2003 code. The calculation show that the UO2 pebble fuel design with 10% enrichment of U235 and 90 ppm burnable poison of Gd2O3 results in the effective multiplication factor (keff) value of 1,01115 at beginning of life (BOL) and become 1,00588 after 2658 days of reactor operation. The core temperature reactivity coefficient is -3.25900E-05 ∆k/k/K and -1,100615E-04 ∆k/k/K at BOL and end of life (EOL), respectively. The reactor is in compliance with inherent safety characteristics indicated by the value of a negative temperature reactivity coefficient. Keywords: PBMR, fuel design, effective multiplication factor, excess reactivity, temperature reactivity coefficient.
DESAIN TERAS DAN BAHAN BAKAR PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR) DENGAN MENGGUNAKAN PROGRAM SRAC Sungkowo Wahyu Santoso; Andang Widiharto; Yohannes Sardjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 16, No 2 (2014): Juni 2014
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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Abstract

Analisis desain down scale teras dan bahan bakar PBMR-HTR dengan menggunakan program SRAC bertujuan mengetahui pengaruh variasi pengayaan U235, burnable poison, laju aliran pendingin dan suhu pendingin masuk terhadap kekritisan teras serta aspek-aspek keselamatan reaktor nuklir dengan parameter nilai keff dan koefisien reaktivitas suhu bahan bakar, moderator dan pendingin. Teras PBMR-HTR berbentuk silinder finite dengan lubang ditengahnya yang berisi 334.000 bahan bakar pebble bed. Bahan bakar berupa UO2, moderator grafit dan pendingin helium. Model desain down scale dilakukan pada ½ teras yang mewakili keseluruhan teras. Penelitian dilakukan dengan memvariasikan pengayaan bahan bakar sebesar 8%, 8,5%, 9%, 9,5% dan 10% sementara variasi konsentrasi burnable poison sebesar 5 ppm, 7 ppm, 9 ppm, 11 ppm, dan 15 ppm. Variasi laju aliran pendingin sebesar 60%, 80%, 100%, 120%, dan 140% sementara variasi suhu masukan pendingin sebesar 673,15K; 723,15K; 773,15K; 823,15K dan 873,15K. Pada penelitian ini keff pada BOL tanpa Gd2O3 sebesar 1.026213 dan EOL sebesar 0.995865 dengan excess reactivity sebesar 2,5 % dengan pengkayaan U235 9%. Sementara keffpada BOL dengan menggunakan Gd2O3 sebesar 1.0069680 dan EOL sebesar 0.9961928 dengan excess reactivity sebesar 0.69 % dengan konsentrasi Gd2O3 7 ppm. Koefisien reaktivitas suhu bahan bakar,moderator dan pendingin berturut-turut sebesar -9,074583E-05/K, -2,971833E-05/K dan 1,120700E-05/K. Koefisien reaktivitas bernilai negatif menunjukkan karakteristik keselamatan melekat (inherent safety) telah terpenuhi. Peningkatan suhu masukan dan penurunan laju aliran pendingin berkontribusi menurunkan nilai keff teras sehingga koefisien reaktivitas bernilai negatif.Kata kunci : PBMR-HTR, kritikalitas, reaktivitas, down scale, burnable poison  Core and fuel down scale analysis on PBMR-HTR using SRAC program aims to identify the influence of U235 enrichment, burnable poison, coolant flow rate and coolant temperature entered to criticality core and safety aspects of nuclear reactor with the parameters are multiplication factor (keff) and fuel temperature coefficient, moderator temperature coefficient and coolant temperature coefficient. Core PBMR-HTR finite cylindrical with a hole in the middle which contains 334,000 pebble fuel bed. That consist of UO2 fuel, graphite moderator and helium coolant. Down scale the design model performed on the half core represent the whole core. The study was conducted by varying the fuel enrichment of 8%; 8.5%; 9%; 9.5% and 10%, while variation burnable poison enrichment at 5 ppm, 7 ppm, 9 ppm, 11 ppm and 15 ppm. The variation of coolant flow rate of 60%, 80%, 100%, 120% and 140% from its original value at 17.118 kg/s while the variation of coolant temperature input at 673.15 K; 723.15 K; 773.15 K; 823.15 K and 873.15 K. In this research, value of keff without Gd2O3 are 1.026213 (BOL) and 1.004173 (EOL) with excess reactivity of 2.55% with 9% U235 enrichment. While keff on BOL by using 7 ppm Gd2O3 of 1.006968 and 1.004198 for EOL with excess reactivity of 0.69%. Fuel temperature reactivity coefficient, moderator and coolant in a row for -8.597317E-05/K; -2.595284E-05 /K and 1.1496E-06/K. Temperature reactivity coefficient is negative. This indicates inherent safety characteristic have been met. Increasing the input temperature and coolant flow rate reduction lowers the value of keff core, and it will contribute to negative reactivity coefficient. Keywords : PBMR-HTR, criticality, reactivity, down scale, burnable poison
A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY Nina Fauziah; Andang Widiharto; Yohannes Sardjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 15, No 2 (2013): Juni 2013
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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Abstract

Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO 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. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA). All materials used were varied in size, according to the value of mean free path for each material. MCNP 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, 2 cm thick of Bi as γ-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. 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. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1.Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk uji in vitro dan in vivo pada Boron Neutron Capture Therapy (BNCT) di Reaktor Riset Kartini dengan menggunakan program Monte Carlo N-Particle (MCNP). Reaktor pada daya sebesar 100 kW digunakan sebagai sumber neutron. Kriteria desain berdasar pada rekomendasi dari IAEA. Setiap material divariasikan ukurannya berdasarkan mean free path radiasi di dalam material tersebut. Simulasi MCNP menunjukkan bahwa dengan menggunakan 5 cm Ni sebagai dinding kolimator, 60 cm Al sebagai moderator, 15 cm 60 Ni sebagai filter, 2 cm Bi sebagai perisai sinar-γ, 3 cm 6Li2CO3-polietilen sebagai penahan radiasi neutron, pada variasi bukaan sebesar 1 sampai 5 cm, dihasilkan fluks neutron epitermalmaksimum sebesar 7,65 x 108 n.cm-2.s-1. Radiasi neutron epitermal tersebut memiliki komponen neutron cepat sebesar 1,76 x 10-13 Gy.cm2.n-1, komponen sinar-γ sebesar1,32 x 10-13 Gy.cm2.n-1, rasio neutron termal per netron epitermal sebesar 0,008, dan direksionalitas maksimum sebesar 0,73. Hasil ini masih tidak memenuhi seluruh kriteria IAEA, karena fluks netron epitermal kurang dari 1,0 x 109 n.cm-2.s-1. Meski demikian, radiasi netron epitermal tersebut masih dapat digunakan karena fluksnya melebihi 5,0 x 108 n.cm-2.s-1. Pada saat diasumsikan bahwa bagian kolom termal yang tersisa di luar daerah kolimator tetap berisi grafit seperti semula, hasil keluaran kolimator menjadi lebih baik dengan fluks neutron maksimum mencapai 1,68 x 109 n.cm-2.s-1. Kata kunci : desain, kolimator, radiasi neutron epitermal, BNCT, MCNP, kriteria
Co-Authors A Suparmi, A Adrian Tesalonika Ahmad Ridwan Angga Dwi Saputra Arthanto, Dwi Handoko Atinna Paramita Gunawan Aulia Setyo Wicaksono Awang Noor Indra Wardana Barlian Gumay Limbaran Beny Yudha Kencana Cahyo Ridho Prabudi Dwijayanto, R. Andika Putra Faridah Faridah Galih Setiaji Gede Sutisna Wijaya Gede Sutrisna Wijaya Hafiz Fahrurrozi Hardiyanti, Iza Shafera Hardy F. Logiani, Hardy F. Irpan Saripudin, Irpan Isman Mulyadi Triatmoko Isman Mulyadi Triatmoko Jans P B Siburian Kusnanto Kusnanto Kutut Suryopratomo M, Irfan Fahmuddin M. Ibnu Khaldun M. Rizaal Martinus I Made Adrian Dwiputra Miftasani, Fitria Nina Fauziah Nur Endah Sari Prayoga Isyan Ranti Warfi Rendra Wahyudityo Sigit Winarto Sihana Sihana Sihana Sihana Slamet Parmanto Sungkowo Wahyu Santoso Taufik Taufik Wijono Wijono Yohannes Sardjono Yohannes Sardjono Yohannes Sardjono Yohannes Sardjono Yohannes Sardjono Yohannes Sardjono