MONTE CARLO N PARTICLE EXTENDED (MCNPX) RADIATION SHIELD MODELLING ON BORON NEUTRON CAPTURE THERAPY FACILITY USING D-D NEUTRON GENERATOR 2.4 MeV

Based Studies were carried out to analyze the internal dose of radiation for workers at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and a room that was actually designed before. This internal dose analyzation included interaction between neutrons and air. The air contained N2 (72%), O2 (20%), Ar (0.93%), CO2, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be below the dose limit for radiation workers which is an amount of 20 mSv/years. From the particles 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 Neutron flux in the air 3.16x107 Neutron/cm2s. The room design in the cancer facility has a measurement of 200 cm in length, 200 cm in width, and 166.40 cm in height. Neutron flux can be used to calculate the reaction rate which is 80.1x10-2 reaction/cm3s for carbon-14 and 8.75x10-5 reaction/cm3s. The internal dose exposed to the radiation worker is 9.08E-9 µSv

Desain Kolimator Tipe Tabung Untuk Penyediaan Berkas Radiografi Dengan Sumber Generator Netron

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

MODELING THE RADIATION SHIELDING OF BORON NEUTRON CAPTURE THERAPY BASED ON 2.4 MEV D-D NEUTRON GENERATOR FACILITY

Radiation shield at Boron Neutron Capture Therapy (BNCT) facility based on D-D Neutron Generator 2.4 MeV has been modified with pre-designed beam shaping assembly (BSA). Modeling includes the material and thickness used in the radiation shield. This radiation shield is expected to protect workers from radiation doses rate that is not exceed 20 mSv·year-1 of dose limit values. The selected materials are barite, paraffin, polyethylene and lead. Calculations were performed using the MCNPX program with tally F4 to determine the dose rate coming out of the radiation shield not exceeding the radiation dose rate of 10 μSv·hr-1. Design 3 was chosen as the recommended model of the four models that have been made. The 3rd shield design uses a 100 cm thickness of barite concrete as primamary layer to surrounding 100 cm x 100 cm x 166.4 cm room, and a 40 cm borated polyethylene surrounding the barite concrete material. Then 10 cm barite concrete and 10 cm of borated polyethylene are added to reduce the primary radiation straight from the BSA after leaving the main layer. The largest dose rate was 4.58 μSv·h-1 on cell 227 and average radiation dose rate 0.65 μSv·hr-1. The dose rates are lower than the lethal dose that is allowed by BAPETEN for radiation worker lethal dose.Keywords: Radiation shield, tally, radiation dose rate, BSA, BNCT PEMODELAN PERISAI RADIASI PADA FASILITAS BORON NEUTRON CAPTURE THERAPY BERBASIS GENERATOR NEUTRON D-D 2,4 MeV. Telah dimodelkan perisai radiasi pada fasilitas Boron Neutron Capture Therapy (BNCT) berbasis reaksi D-D pada Neutron Generator 2,4 MeV dengan Beam Shaping Assembly (BSA) yang telah didesain sebelumnya. Pemodelan ini dilakukan untuk memperoleh suatu desain perisai radiasi untuk fasilitas BNCT berbasis generator neutron 2,4 MeV. Pemodelan dilakukan dengan cara memvariasikan bahan dan ketebalan perisasi radiasi. Bahan yang dipilih adalah beton barit, parafin, polietilen terborasi dan timbal. Perhitungan dilakukan menggunakan program MCNPX dengan tally F4 untuk menentukan laju dosis yang keluar dari perisai radiasi. Desain periasi radiasi dinyatakan optimal jika radiasi yang dihasilkan diluar perisai radiasi tidak melebihi Nilai Batas Dosis (NBD) yang telah ditentukan oleh BAPETEN. Hasilnya, diperoleh suatu desain perisai radiasi menggunakan lapisan utama beton barit setebal 100 cm yang mengelilingi ruangan 100 cm x 100 cm x 166,4 cm dan polietilen terborasi 40 cm yang mengelilingi bahan beton barit. Kemudian ditambahkan beton barit 10 cm dan polietilen terborasi 10 cm untuk mengurangi radiasi primer yang lurus dari BSA setelah keluar dari lapisan utama. Laju dosis terbesar adalah 4,58 μSv·jam-1 pada sel 227 dan laju dosis rata-rata yang dihasilkan adalah sebesar 0,65 µSv·jam-1. Nilai laju dosis tersebut masih dibawah ambang batas NBD yang diperbolehkan oleh BAPETEN untuk pekerja radiasi.Kata kunci: Perisai radiasi, tally, laju dosis radiasi, BSA, BNC

Assesment of Boron Neutron Capture Therapy (BNCT): Compact Neutron Generators

Boron  Neutron Capture Therapy (BNCT)  is an effective and promising treatment          of tumour types which are resistant to conventional therapies. The characteristics of boron neutron capture therapy (BNCT) for cancer treatment demand, in addition to sufficient fluxes of epithermal neutrons, proper conditions of the neutron sources—compact layout, flexible operation, compatibility with hospital setting, etc. The lack of proper neutron sources that can be integrable to the infrastructure of hospital or clinical facilities is a major problem. Compact neutron generators (CNGs), which are the most compact and least expensive, were a potential, alternative, solution to existing BNCT treatment facilities based on nuclear reactors. This paper will provide information about the latest CNGs technology development that has contributed to the Boron Neutron Capture Therapy (BNCT) technology improvement

OPTIMIZATION OF A NEUTRON BEAM SHAPING ASSEMBLY DESIGN FOR BNCT AND ITS DOSIMETRY SIMULATION BASED ON MCNPX

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, MCNP

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

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

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

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

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

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, kriteri

Design Collimator and Dosimetry of in Vitro and in Vivo Test Using MCNP-X Code

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

RADIOACTIVE WASTE TREATMENT OF OIL AND GAS INDUSTRY

The increase of energy needs leads to the increase of oil and gas industry in Indonesia. Along with this, a good understanding about the application of radioactive materials, the treatment of radioactive waste and its disposal is needed. It is important because nowadays the radiation-based technologies are widely used by oil and gas industry, and a good understanding of the mentioned points will enable us to provide better and improved treatment as we strive to achieve clean technology. The uses of radioactive materials in the oil and gas industry are in exploration, producing, refineries, inspection of facilities, laboratories, and industrial security. The treatment of radioactive waste from oil and gas industry are divided into aqueous waste treatment, organic liquid waste treatment, and solid waste treatment