Garuda - Garba Rujukan Digital

p-Index From 2019 - 2024

0.659

P-Index

This Author published in this journals

All Journal Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega Ganendra: Majalah IPTEK Nuklir Jurnal Pengembangan Energi Nuklir Indonesian Journal of Physics and Nuclear Applications

Yohannes Sardjono

Unknown Affiliation

Author-ID : 2656758

Computer Science & IT Decision Sciences, Operations Research & Management Earth & Planetary Sciences Electrical & Electronics Engineering Energy Engineering Environmental Science Industrial & Manufacturing Engineering Materials Science & Nanotechnology Physics Social Sciences Other

Published : 9 Documents Claim Missing Document

Claim Missing Document

Articles

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.

Distribution of Water Phantom BNCT Kartini Research Reactor Based Using PHITS Nunung Gupita Ratnasari; Susilo Susilo; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 3 No 2 (2018)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

The purpose of this research was to calculate the radiation dose on BNCT. Boron Neutron Capture Therapy (BNCT) is a cancer therapy which utilizes thermal neutron-capture reactions by boron-10 isotopes that produce alpha particles and lithium nuclei. The advantage of BNCT is that radiation effects can be limited to tumor cells. The dose of radiation on BNCT depends heavily on the distribution of boron and the neutron free region. The calculation method involves alpha and lithium particles of reactions having high Linear Energy Transfer (LET). By replacing the target of using water phantom that contains heavy water and covered by acrylic glass measuring 30 cm x 30 cm x 30 cm, the dose is calculated using PHITS-based applications. By comparing the simulation results between boron and phantom water or phantom without boron then the conclusion is the absorbed dose of phantom water containing boron is larger than phantom water without boron.

THE DOSE ANALYSIS OF BORON NEUTRON CAPTURE THERAPY (BNCT) TO THE BRAIN CANCER (GLIOBLASTOMA MULTIFORM) USING MCNPX-CODE WITH NEUTRON SOURCE FROM COLLIMATED THERMAL COLUMN KARTINI RESEARCH NUCLEAR Kholidah Hasyim; Yohannes Sardjono; Yosaphat Sumardi
Indonesian Journal of Physics and Nuclear Applications Vol 3 No 3 (2018)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

This research was aimed at discovering the optimum concentration of Boron-10 in concentrations range 20 µgram/gram until 35 µgram/gram with Boron Neutron Capture Therapy (BNCT) methods and the shortest time irradiation for cancer therapy. The research about dose analysis of Boron Neutron Capture Therapy (BNCT) to the brain cancer (Glioblastoma Multiform) using MCNPX-Code with a neutron source from Collimated Thermal Column Kartini Research Nuclear has been conducted. This research was a simulation-based experiment using MCNPX, and the data was arranged on a graph using OriginPro 8. The modelling was performed with the brain that contains cancer tissue as a target and the reactor as a radiation source. The variations of Boron concentrations in this research was on 20, 25, 30 and 35 μg/gram tumours. The outputs of MCNP were neutron scattering dose, gamma ray dose and neutron flux from the reactor. Neutron flux was used to calculate the doses of alpha, proton and gamma rays produced by the interaction of tissue material and thermal neutrons. Based on the calculations, the optimum concentration of Boron-10 in tumour tissue was for a 30 µg/gram tumour with the radiation dose in skin at less than 3 Gy. The irradiation times required were 2.79 hours for concentration 20 μg/gram ; 2.78 hours for concentration 25 μg/gram ; 2.77 hours for concentration 30 μg/gram ; 2.8 hours for concentration 35 μg/gram.

Distribution of Water Phantom BNCT Cyclotron based Using PHITS siti maimanah; siti maimanah; Susilo Susilo; Yohannes Sardjono
Indonesian Journal of Physics and Nuclear Applications Vol 4 No 1 (2019)
Publisher : Fakultas Sains dan Matematika Universitas Kristen Satya Wacana

This research purpose is to estimate the dose distribution of BNCT in water phantom. Some common methods in the treatment of cancer such as brakhiterapi, surgery, chemotherapy, and radiotherapy still have the risk of damaging healthy tissue around cancer cells. BNCT is a selectively-designed technique by targeting high-loaded LET particles to tumors at the cellular level. BNCT proves to be a powerful method of killing cancer without damaging normal tissue. The source of the neutron used from the cyclotron dose in water phantom with the size of 30 cm x 30 cm x 30 cm was calculated using PHITS program. The result from the simulation is that boron water panthom has a dosimetri higher than phantom water without boron.

RADIATION DOSE OPTIMIZATION OF BREAST CANCER WITH PROTON THERAPY METHOD USING PARTICLE AND HEAVY ION TRANSPORT CODE SYSTEM Milah Fadhilah Kusuma Fasihu; Andang Widi Harto; Isman Mulyadi Triatmoko; Gede Sutrisna Wijaya; Yohannes Sardjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6290

Radiotherapy is one of the cancer treatments conducted by giving a high dose to the tumor target and minimizing the dose exposed in the healthy organs. One of the methods is proton therapy. Proton therapy is usually used in several breast cancer cases by minimizing the damage in the surrounding tissues due to having good precision. In this study, proton therapy in breast cancer will be simulated. This study aims to identify the optimal dose in breast cancer therapy using proton therapy and to identify the dose exposed in the healthy organs surrounding cancer. This study is PHITS program simulation-based to model the geometry and the components of breast cancer and the surrounding organs. The source of radiation used is proton which is the output of proton therapy with proton/sec firing intensity. The variation in beam modelling towards the dose profile of the tumor used is uniform and pencil beam. The proton energy used is 70 MeV up to 120 MeV. The result of this study shows that the dose from using pencil beam scanning technic of proton therapy for breast cancer is 50.3997 Gy (W) with the total amount of fraction is 25 and the result of dose below the threshold dose in the healthy organs is the skin gets 4.4.0553 Gy per fraction, the left breast gets 0,0011 Gy per fraction, the right breast gets 2.6469 Gy per fractions, the right lung gets 0.0125 Gy per fraction, the left lung gets 0.029 Gy per fraction, the rib gets 0.0179 Gy per fraction, and the heart gets 0.0077 Gy per fraction.

DOSE DISTRIBUTION ANALYSIS OF PROTON THERAPY FOR MEDULLOBLASTOMA CANCER WITH PHITS 3.24 Moh. Miftakhul Dwi Fianto; Yohannes Sardjono; Andang Widi Harto; Isman Mulyadi Triatmoko; Gede Sutresna Wijaya; Yaser Kasesaz
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6581

One of the developments in particle therapy is proton radiation therapy. Meanwhile, a limited number of available proton therapy facilities makes research related to proton therapy difficult. Therefore, there is a need for alternative proton therapy simulations using programs other than those in proton therapy facilities. This research was aimed to simulate medulloblastoma brain cancer which children often experience.The program used in this research was PHITS version 3.24. The human body was modeled with the revised ORNL-MIRD phantom for a 10-year-old child. The therapy scheme was a whole posterior fossa boost of 19.8 Gy. The proton passive scattering was simulated by passing a uniform proton beam through the aperture and compensator with energy variations. The proton pencil beam scanning was simulated with small cylindrical beams with a radius of 0.5 cm, which were adjusted to the planning target volume with layers variations.The total duration to give the prescription dose was 550 seconds with passive scattering and 605 seconds with pencil beam scanning. In passive scattering, the OAR(s) with the most significant percentage of absorbed dose were the skin, cranium, and muscle, i.e., 8.22 ± 0.15 %, 5.51 ± 0.05 % and 1.39 ± 0,04 % respectively to their maximum tolerated dose, while in the pencil beam scanning, the OAR(s) with the most significant percentage of absorbed dose were the skin, cranium, and muscle, i.e., 5.42 ± 0.08 %, 4.43 ± 0.05 % and 0.51 ± 0.05 % respectively to their maximum tolerated dose. Dose distribution in passive scattering was relatively better than in pencil beam scanning in terms of dose homogeneity using dose sampling analysis at some points within the planning target volume.

ANALISIS KESELAMATAN REAKTOR KARTINI BERDASAR KEJADIAN PEMICU YANG DIPOSTULASIKAN Yohannes Sardjono; Eko Priyono; Syarip Syarip
GANENDRA Majalah IPTEK Nuklir Volume 8 Nomor 2 Juli 2005
Publisher : Website

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.

DESAIN KOLIMATOR TIPE TABUNG UNTUK PENYEDIAAN BERKAS RADIOGRAFI DENGAN SUMBER GENERATOR NETRON Yohannes Sardjono
GANENDRA Majalah IPTEK Nuklir Volume 10 Nomor 2 Juli 2007
Publisher : Website

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.

DESAIN TERAS DAN BAHAN BAKAR PLTN JENIS HTR-PBMR PADA DAYA 50 MWe DENGAN MENGGUNAKAN PROGRAM SRAC2006 Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono
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 | DOI: 10.17146/jpen.2014.16.1.2559

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.

Co-Authors Abdullah Nur Aziz Andang Widi Harto Aniti Payudan Bima Caraka Putra Eko Priyono Gede Sutresna Wijaya Gede Sutrisna Wijaya Isman Mulyadi Triatmoko Kholidah Hasyim Milah Fadhilah Kusuma Fasihu Moh. Miftakhul Dwi Fianto Nunung Gupita Ratnasari siti maimanah siti maimanah Susilo Susilo Susilo Susilo Syarip Syarip Yaser Kasesaz Yosaphat Sumardi

Title

Found 9 Documents
Search

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Title Search

Found 9 Documents Search

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Title

Found 9 Documents
Search