Articles

DESAIN KOLIMATOR TIPE TABUNG UNTUK PENYEDIAAN BERKAS RADIOGRAFI DENGAN SUMBER GENERATOR NETRON Sardjono, Yohannes
Jurnal Iptek Nuklir Ganendra Vol 10, No 2 (2007)
Publisher : BATAN

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Abstract

DESAIN KOLIMATOR TIPE TABUNG UNTUK PENYEDIAAN BERKAS RADIOGRAFI DENGAN SUMBERGENERATOR NETRON. Telah dilakukan desain kolimator untuk penyediaan berkas radiografi netron dengansumber generator netron. Kolimator ini berguna untuk mendapatkan fluks netron termal yang optimal dengan pengotorradiasi (netron epitermal dan gamma) yang sekecil-kecilnya. Proses desain dilakukan dengan melakukan simulasimenggunakan Monte Carlo N-Particle (MCNP) code untuk menghitung tally berupa fluks netron dan laju dosisekuivalen. Desain kolimator yang dipilih adalah jenis tabung yang tersusun dari material moderator parafin, reflektorgrafit, dan kolimator wall alumunium. Parameter optimasi desain adalah panjang kolimator 4 - 8 cm, dengan interval 1cm, jenis bahan moderator (parafin, grafit, berilium, dan air), jenis beam filter adalah timbal, dan material apertureadalah boron atau kadmium. Kriteria penerimaan adalah fluks netron termal 103 - 106 n.cm-2.s-1, n/γ ratio > 106n.cm-2.mR-1 dan Cd ratio > 2. Untuk keselamatan lingkungan digunakan parafin sebagai biological shielding dan timbalsebagai casing. Dari hasil perhitungan optimasi desain dapat diperoleh bahwa kolimator dengan sumber generatornuetron menghasilkan keluaran fluks netron termal 4.67 + 0.5981 x 103 n.cm-2.s-1, rasio netron-gamma (n/γ) ≥ (1.56 +0,000111).106 n.cm–2 mR-1 dan laju dosis ekuivalen pada jarak 10 cm dari permukaan fasilitas adalah 0,0378 - 0,0521mR/jam.
ANALISIS KESELAMATAN REAKTOR KARTINI BERDASAR KEJADIAN PEMICU YANG DIPOSTULASIKAN Sardjono, Yohannes; Priyono, Eko; Syarip, Syarip
Jurnal Iptek Nuklir Ganendra Vol 8, No 2 (2005)
Publisher : BATAN

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ANALISIS KESELAMATAN REAKTOR KARTINI BERDASAR KEJADIAN PEMICU YANG DIPOSTULASIKAN.Berdasarkan analisis kejadian pemicu yang dipostulasikan maka ada 8 kejadian yang dipostulasikan (PostulatedInitiating Event) : seperti kehilangan catu daya listrik, kegagalan sistem scram, kehilangan aliran pendingin,kehilangan pendingin, kegagalan transfer cask, kejadian internal/eksternal dan kesalahan manusia. Dari 8 kejadiantersebut, hanya satu kejadian yang menyebabkan terlepasnya bahan radioaktif dari seluruh sistem bahan bakar kelingkungan yaitu kejadian gagalnya sistem pemindah bahan bakar (transfer cask). Urutan kejadiannya adalahtransfer cask jatuh di atas teras reaktor dan mengakibatkan seluruh kelongsong bahan bakar pecah lalu diikutidengan hilangya seluruh air tangki reaktor sehingga seluruh inti hasil belah gas yang ada di celah bahan bakar lepaske lingkungan. Analisis terlepasnya bahan radioaktif ke lingkungan menggunakan paket program dengan bahasaTurbo Pascal dan lama eksekusi 5 menit. Dari hasil analisis diperoleh bahwa dosis radiasi gamma yang diterimaoleh penduduk pada saat 2 jam setelah terjadi kecelakaan pada radius 33 meter adalah 25 rem dan dosis iodinadalah 300 rem berarti proses evakuasi sangat sederhana karena tidak melibatkan penduduk di sekitar kawasanP3TM.
OPTIMIZATION OF A NEUTRON BEAM SHAPING ASSEMBLY DESIGN FOR BNCT AND ITS DOSIMETRY SIMULATION BASED ON MCNPX Ardana, I Made; Sardjono, Yohannes
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 19, No 3 (2017): Oktober 2017
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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Abstract

This article involves two main objectives of BNCT system. The first goal includes optimization of 30 MeV Cyclotron-based Boron Neutron Capture Therapy (BNCT) beam shaping assembly. The second goal is to calculate the neutron flux and dosimetry system of BNCT in the head and neck soft tissue sarcoma. A series of simulations has been carried out using a Monte Carlo N Particle X program to find out the final composition and configuration of a beam shaping assembly design to moderate the fast neutron flux, which is generated from the thick beryllium target. The final configuration of the beam shaping assembly design includes a 39 cm aluminum moderator, 8.2 cm of lithium fluoride as a fast neutron filter and a 0.5 cm boron carbide as a thermal neutron filter. Bismuth, lead fluoride, and lead were chosen as the aperture, reflector, and gamma shielding, respectively. Epithermal neutron fluxes in the suggested design were 2.83 x 109 n/s cm-2, while other IAEA parameters for BNCT beam shaping assembly design have been satisfied. In the next step, its dosimetry for head and neck soft tissue sarcoma is simulated by varying the concentration of boron compounds in ORNL neck phantom model to obtain the optimal dosimetry results. MCNPX calculation showed that the optimal depth for thermal neutrons was 4.8 cm in tissue phantom with the maximum dose rate found in the GTV on each boron concentration variation. The irradiation time needed for this therapy were less than an hour for each level of boron concentration.Keywords: Optimization, Beam Shaping Assembly, BNCT, Dosimetry, 30 MeV Cyclotron, MCNPX. OPTIMASI DESAIN KOLIMATOR NEUTRON UNTUK SISTEM BNCT DAN UJI DOSIMETRINYA MENGGUNAKAN PROGRAM MCNPX. Telah dilakukan penelitian tentang sistem BNCT yang meliputi dua tahapan simulasi dengan menggunakan program MCNPX yaitu uji simulasi untuk optimasi desain kolimator neutron untuk sistem BNCT berbasis Siklotron 30 MeV dan uji simulasi untuk menghitung fluks neutron dan dosimetri radiasi pada kanker sarkoma jaringan lunak pada leher dan kepala. Tujuan simulasi untuk mendapatkan desain kolimator yang paling optimal dalam memoderasi fluks neutron cepat yang dihasilkan dari sistem target berilium sehingga dapat dihasilkan fluks neutron yang sesuai untuk sistem BNCT. Uji optimasi dilakukan dengan cara memvariasikan bahan dan ketebalan masing-masing komponen dalam kolimator seperi reflektor, moderator, filter neutron cepat, filter neutron thermal, filter radiasi gamma dan lubang keluaran. Desain kolimator yang diperoleh dari hasil optimasi tersusun atas moderator berbahan Al dengan ketebalan 39 cm, filter neutron cepat berbahan LiF2 setebal 8,2 cm, dan filter neutron thermal berbahan B4C setebal 0,5 cm. Untuk reflektor, filter radiasi gamma dan lubang keluaran masing-masing menggunakan bahan PbF2, Pb dan Bi. Fluks neutron epithermal yang dihasilkan dari kolimator yang didesain adalah sebesar 2,83 x 109 n/s cm-2 dan telah memenuhi seluruh parameter fluks neutron yang sesuai untuk sistem BNCT. Selanjutnya uji simulasi dosimetri pada kanker sarkoma jaringan lunak pada leher dan kepala dilakukan dengan cara memvariasikan konsentrasi senyawa boron pada model phantom leher manusia (ORNL). Selanjutnya model phantom tersebut diiradiasi dengan fluks neutron yang berasal dari kolimator yang telah didesain sebelumnya. Hasilnya, fluks neutron thermal mencapai nilai tertinggi pada kedalaman 4,8 cm di dalam model phantom leher ORNL dengan laju dosis tertinggi terletak pada area jaringan kanker. Untuk masing-masing variasi konsentrasi senyawa boron pada model phantom leher ORNL supaya dapat mematikan jaringan kanker, membutukan waktu iradiasi neutron kurang dari satu jam.Kata kunci: Optimasi, Kolimator, BNCT, Dosimetri, Siklotron 30 MeV, MCNPX
STUDI DESAIN DOWN SCALE TERAS REAKTOR DAN BAHAN BAKAR PLTN JENIS PEBBLE BED MODULAR REACTOR – HTR 100 MWe Parmanto, Slamet; Widiharto, Andang; Sardjono, Yohannes
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

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.
PEMODELAN KOLIMATOR DI RADIAL BEAM PORT REAKTOR KARTINI UNTUK BORON NEUTRON CAPTURE THERAPY Vallenry, Bemby Yulio; Widiharto, Andang; Sardjono, Yohannes
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 16, No 1 (2014): Pebruari 2014
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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Abstract

Salah satu metode terapi kanker adalah Boron Neutron Capture Therapy (BNCT). BNCT memanfaatkan tangkapan neutron oleh 10B yang terendapkan pada sel kanker. Keunggulan BNCT dibandingkan dengan terapi radiasi lainnya adalah tingkat selektivitas yang tinggi karena tingkatannya adalah sel. Pada penelitian ini dilakukan pemodelan kolimator di radial beamport reaktor Kartini sebagai dasar pemilihan material dan manufature kolimator sebagai sumber neutron untuk BNCT. Pemodelan ini dilakukan dengan simulasi menggunakan perangkat lunak Monte Carlo N-Particle versi 5 (MCNP 5). MCNP 5 adalah suatu paket program untuk memodelkan sekaligus menghitung masalah transpor partikel dengan mengikuti sejarah hidup neutron semenjak lahir, bertranspor pada bahan hingga akhirnya hilang karena mengalami reaksi penyerapan atau keluar dari sistem. Pemodelan ini menggunakan variasi material dan ukurannya agar menghasilkan nilai dari tiap parameter-parameter yang sesuai dengan rekomendasi I International Atomic Energy Agency (IAEA) untuk BNCT, yaitu fluks neutron epitermal (Фepi) > 9 n.cm-2.s-1, rasio antara laju dosis neutron cepat dan fluks neutron epitermal (Ḋf/Фepi) < 2,0 x 10-13 Gy.cm2.n-1, rasio antara laju dosis gamma dan fluks neutron epitermal (Ḋγ/Фepi) < 2,0 x 10-13 Gy.cm2.n-1, rasio antara fluks neutron termal dan epithermal (Фth/Фepi) < 0,05 dan rasio antara arus dan fluks neutron epitermal (J/Фepi) > 0,7. Berdasarkan hasil optimasi dari pemodelan ini, material dan ukuran penyusun kolimator yang didapatkan yaitu 0,75 cm Ni sebagai dinding kolimator, 22 cm Al sebagai moderator dan 4,5 cm Bi sebagai perisai gamma. Keluaran berkas radiasi yang dihasilkan dari pemodelan kolimator radial beamport yaitu Фepi = 5,25 x 106 n.cm-2s-1, Ḋf/Фepi =1,17 x 10-13 Gy.cm2.n-1, Ḋγ/Фepi = 1,70 x 10-12 Gy.cm2.n-1, Фth/Фepi = 1,51 dan J/Фepi = 0,731. Berdasarkan penelitian ini, hasil optimasi 5 parameter sebagai persyaratan kolimator untuk BNCT yang keluar dari radial beam port tidak sepenuhnya memenuhi kriteria yang direkomendasikan oleh IAEA sehingga perlu dilakukan penelitian lebih lanjut agar tercapainya persyaratan IAEA.Kata kunci: BNCT, radial beamport, MCNP 5, kolimator  One of the cancer therapy methods is BNCT (Boron Neutron Capture Therapy). BNCT utilizes neutron nature by 10B deposited on cancer cells. The superiority of BNCT compared to the rradiation therapy is the high level of selectivity since its level is within cell. This study was carried out on collimator modelling in radial beam port of reactor Kartini for BNCT. The modelling was conducted by simulation using software of Monte CarloN-Particle version5 (MCNP 5). MCNP5 is a package of the programs for both simulating and calculating the problem of particle transport by following the life cycle of a neutron since its birth from fission reaction, transport on materials, until eventually lost due to the absorption reaction or out from the system. The collimator modelling used materials which varied in size in order to generate the value of each of the parameters in accordance with the recommendation of the IAEA, the epithermal neutron flux (Фepi) > 1.0 x 109n.cm-2s-1, the ratio between the neutron dose rate fast and epithermal neutron flux (Ḋf/Фepi) < 2.0 x10-13 Gy.cm2.n-1, the ratio of gamma dose rate and epithermal neutron flux (Ḋγ/Фepi) < 2.0 x10-13 Gy.cm2.n-1, the ratio between the thermal and epithermal neutron flux (Фth/Фepi) < 0.05 and the ratio between the current and flux of the epithermal neutron (J/Фepi) > 0.7. Based on the results of the optimization of the modeling, the materials and sizes of the collimator construction obtained were 0.75 cm Ni as collimator wall, 22 cm Al as a moderator and 4.5 cm Bi as a gamma shield. The outputs of the radiation beam generated from collimator modeling of the radial beam port were Фepi = 5.25 x 106 n.cm-2.s-1, Ḋf/Фepi = 1.17 x 10-13 Gy.cm2.n-1, Ḋγ/Фepi = 1.70 x 10-12 Gy.cm2.n-1, Фth/Фepi = 1.51 and J/Фepi = 0.731. Based on this study, the results of the beam radiation coming out of the radial beam port did not fully meet the criteria recommended by the IAEA so need to continue this study to get the criteria of IAEA. Keywords: BNCT, radial beamport, MCNP 5, collimator
DESAIN TERAS DAN BAHAN BAKAR PLTN JENIS HTR-PBMR PADA DAYA 50 MWe DENGAN MENGGUNAKAN PROGRAM SRAC2006 Caraka Putra, Bima; Sumardi, Yosaphat; Sardjono, Yohannes
Jurnal Pengembangan Energi Nuklir Vol 16, No 1 (2014): Juni 2014
Publisher : Pusat Kajian Sistem Energi Nuklir, Badan Tenaga Nuklir Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (286.82 KB) | DOI: 10.17146/jpen.2014.16.1.2559

Abstract

ABSTRAK DESAIN TERAS DAN BAHAN BAKAR PLTN JENIS HTR-PBMR PADA DAYA 50 MWe DENGAN MENGGUNAKAN PROGRAM SRAC2006. Penelitian ini bertujuan untuk mengkaji desain teras dan bahan bakar PLTN jenis HTR-PBMR (HIGH TEMPERATURE REACTOR - PEBBLE BED MODULAR REACTOR) 50 MWe dari keadaan Beginning of Life (BOL) sampai Ending of Life (EOL) dengan masa operasi 8 tahun. Parameter yang dianalisis dalam penelitian ini adalah distribusi suhu di dalam teras, persen pengkayaan U235, komposisi bahan bakar, kekritisan, dan koefisien reaktivitas suhu teras. Penelitian dilakukan dengan menyiapkan data parameter desain teras antara lain densitas nuklida, dimensi bahan bakar dan teras, dan distribusi suhu aksial teras. Paket program SRAC2006 digunakan untuk mendapatkan nilai faktor multiplikasi effektif (keff) teras dari data input yang telah disiapkan. Hasil penelitian menunjukkan nilai kekritisan teras berbanding lurus dengan penambahan pengkayaan U235. Pengayaan optimum tanpa penggunaan burnable poison didapatkan pada nilai 10,125% dengan reaktifitas lebih sebesar 3,12% pada BOL. Penambahan burnable poison Gd2O3 didapat nilai optimumnya sebesar 12 ppm dengan nilai reaktifitas lebih pada BOL sebesar 0,38%. Untuk penggunaan Er2O3 nilai optimumnya adalah 290 ppm dengan reaktifitas lebih 1,24% pada saat BOL. Koefisien reaktivitas suhu teras tanpa burnable poison dan penggunaan Gd2O3 dan Er2O3 bernilai negatif yang menunjukkan sifat inherent safety-nya. Kata kunci: desain, teras, bahan bakar, PLTN, SRAC2006. ABSTRACT DESIGN OF 50 MWe HTR-PBMR REACTOR CORE AND NUCLEAR POWER PLANT FUEL USING SRAC2006 PROGRAMME. This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (keff) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.12% at BOL. The addition Gd2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er2O3 with an optimum value 290 ppm has an excess reactivity 1.24% at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. Keywords: design, fuel, nuclear power plant, SRAC2006.
CORE DESIGNS OF ABWR FOR PROPOSED OF THE FIRST NUCLEAR POWER PLANT IN INDONESIA Sardjono, Yohannes; Aritomi, Masanori; Fennern, Larry E.
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 13, No 1 (2011): Pebruari 2011
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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Indonesia as an archipelago has been experiencing high growth industry and energy demand due to high population growth, dynamic economic activities. The total population is around 230 million people and 75 % to the total population is living in Java. The introduction of Nuclear Power Plant on Java Bali electricity grid will be possible in 2022 for 2 GWe, using proven technology reactor like ABWR or others light water reactor with nominal power 1000 MWe. In this case, the rated thermal power for the equilibrium cycles is 3926 MWt, the cycle length is 18 month and overall capacity factor is 87 %. The designs were performed for an 872-fuel bundles ABWR core using GE-11 fuel type in an 9×9 fuel rod arrays with 2 Large Central Water Rods (LCWR). The calculations were divided into two steps; the first is to generate bundle library and the other is to make the thermal and reactivity limits satisfied for the core designs. Toshiba General Electric Bundle lattice Analysis (TGBLA) and PANACEA computer codes were used as designs tools. TGBLA is a General Electric proprietary computer code which is used to generate bundle lattice library for fuel designs. PANACEA is General Electric proprietary computer code which is used as thermal hydraulic and neutronic coupled BWR core simulator. This result of core designs describes reactivity and thermal margins i.e.; Maximum Linear Heat Generation rate (MLHGR) is lower than 14.4 kW/ft, Minimum Critical Power Ratio (MCPR) is upper than 1.25, Hot Excess Reactivity (HOTXS) is upper than 1 %Dk at BOC and 0.8 %Dk at 200 MWD/ST and Cold Shutdown Margin Reactivity (CSDM) is upper than 1 %Dk. It is concluded that the equilibrium core design using GE-11 fuel bundle type satisfies the core design objectives for the proposed of the firs Indonesia ABWR Nuclear Power Plant.Keywords: The first NPP in Indonesia, ABWR-1000 MWe, and core designs.   Indonesia adalah sebagai negara kepulauan yang laju pertumbuhan industri, energi, penduduk dan ekonominya cukup tinggi. Pada saat ini, jumlah penduduk Indonesia ada sekitar 230 juta dan 75 % dari jumlah penduduk tersebut tinggal di Pulau Jawa. Pada tahun 2022, dimungkinkan sistem jaringan Jawa-Bali dapat menerima beban 2 unit PLTN yang teknologinya sudah teruji seperti PLTN ABWR atau PLTN air ringan lainnya yang kapasitasnya masing-masing 1 GW. Untuk itu diambilah contoh perhitungan untuk PLTN ABWR pada siklus keseimbangan dengan daya termal 3926 MWt dan lama operasi 18 bulan dan kapasitas faktornya minimum 87 %. Desain ini telah dicapai dengan jumlah bahan bakar teras 872 bundel bahan bakar tipe GE-11 yang susunannya 9×9 batang bahan bakar yang ditengahnya ditempatkan 2 bahan bakar besar tiruan yang berisi air. Ada 2 langkah perhitungan; pertama adalah menggenerasikan pustaka data bundel bahan bakar dan selanjutnya digunakan untuk analisis termal dan reaktivitas dalam teras. Desain teras menggunakan kode komputer Toshiba General Electric Bundle Lattice Analysis (TGBLA) dan PANACEA. TGBLA adalah sebuah kode komputer yang dimiliki oleh General Electric Nuclear Energy untuk menggenerasikan pustaka data dalam sistem satuan cell dalam setiap batang bahan bakar dalam setiap bundle. PANACEA adalah kode komputer milik General Electric yang digunakan untuk analisis thermal hydraulic dan netronik yang digabung dalam simulator PLTN BWR. Hasil desain teras menguraikan tentang karakteristik termal dan reaktivitas teras seperti; laju maksimum pembangkitan panas linier (MLHGR) adalah lebih rendah dari 14,4 kW/ft, rasio daya kritis minimum (MCPR) adalah diatas dari 1,25, Reaktivitas Panas Lebih (HOTXS) adalah lebih besar dari 1 %Dk pada BOC dan 0,8 %Dk pada 200 MWD/ST dan reaktivitas shutdown margin dingin (CSDM) adalah lebih besar dari 1 %Dk. Untuk itu dapat disimpulkan bahwa desain teras PLTN ABWR pertama untuk diusulkan dibangun pertama di Indonesia dengan menggunakan bundle bahan bakar tipe GE-11 adalah telah memenuhi persyaratan dan tujuan desain. Kata kunci: PLTN pertama di Indonesia, ABWR-1000 MWe, dan desain teras.
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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (1133.98 KB) | DOI: 10.24246/ijpna.v1i1.1-13

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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (601.551 KB) | DOI: 10.24246/ijpna.v1i1.14-19

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

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (602.52 KB) | DOI: 10.24246/ijpna.v1i1.20-33

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.
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 Simangunsong, Deo Clinton Maranatha 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