Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator
Keywords:
radiation shielding, design, BNCT, MCNPX, thermal column
Abstract
Studies were carried out to design a shielding for BNCT facility in the end of Kartini reactor’s thermal column with predesigned collimator. The design consist of selecting the material and their thickness. The shielding is required to absorb the leaking radiation until the Dose Limit Value of 20 mSv/year for radiation worker is met. The material considered were paraffin, barite concrete, borated polyethylene, stainless steel 304 and lead. The calculation was done using MCNPX tally facility with converted dose limit value of 10.42 µSv/hour. Design number two were chosen as the best from three designs which surrounded a room with length, width and height of, respectively 200 cm, 200 cm and 166.4 cm. The first and main layer are borated polyethyelene and barite concrete of 20 and 30 cm, respectively. The additional layer are borated polyethyelene and barite concrete of 15 cm and 15 cm with less volume than the main layer to decrease the primary straight radiation from the thermal column. Maximum radiation dose rate is 7.0746 µ Sv/hour in cell 227 with average dose rate of 2.58712 µSv/hour.Downloads
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References
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Bielajew, A. F. (2001). Fundamentals of The Monte Carlo Method for neutral and charged particle transport. Michigan: The University of Michigan.
Calzada, E. (2011). Reusable shielding material for neutron- and gamma-radiation. Nuclear Instruments and Methods in Physics Research A, 77-80.
Cember, H., & Johson, T. E. (2009). Introduction to Health Physics. Illinois: The McGraw-Hill.
Gencel, O. (2011). Determination and calculation of gamma and neutron shielding characteristics of concretes containing different hematite proportions. Annals of Nuclear Energy, 2719-2723.
Ghiasi, H. (2012). A new analytical formula for neutron capture gamma dosecalculations in double-bend mazes in radiation therapy. Reports of practical oncology and radiotherapy, 220-225.
Indonesia, K. B. (2013). Peraturan Kepala Badan Pengawas Tenaga Nuklir Nomor 4 Tahun 2013 Tentang Proteksi dan Keselamatan Radiasi Dalam Pemanfaatan Tenaga Nuklir. Jakarta: BAPETEN.
Irhas. (2013). Dosimetri Boron Neutron Capture Therapy pada Kanker Hati (Hepatocellular Carinoma) Menggunakan MCNP-Code dengan Sumber Netron dari Kolom Termal Reaktor Kartini. Yogyakarta: Universitas Gadjah Mada.
Lamarsh, J. R. (1983). Introduction to Nuclear Engineering. Massachusetts: Addison-Wesley.
Pelowitz, D. B. (2008). MCNPX User's Manual. New Mexico: Los Alamos National Laboratory.
Pouryavi, M. (2015). Radiation shielding design of BNCT treatment room for D-T neutron source. Applied Radiation and Isotopes, 90-96.
Protection, I. C. (1975). Report of the task group on Reference Man. ICRP Publication 23. Oxford: Pergamon Press.
Santoso, B. H. (2014). Pemodelan Perisai Radiasi Fasilitas BNCT Dengan Sumber Beamport Tembus Teras Reaktor Kartini Menggunakan MCNP5. Yogyakarta: Universitas Gadjah Mada.
Singh, V. P., & Badiger, N. (2014). Gamma ray and neutron shielding properties of some alloy materials. Annals of Nuclear Energy, 301-310.
Solleh, M. R., Tajudiin, A. A., Mohamed, A. A., Eid, E. M., Munem, A., Rabir, M. H., . . . Yoshiaki, K. (2011). COLLIMATOR AND SHIELDING DESIGN FOR BORON NEUTRON CAPTURE THERAPY (BNCT) FACILITY AT TRIIGA MARK II REACTOR. JOURNAL OF NUCLEAR And Related TECHNOOGOIES Volume 8 No.2 .
Stella, S. (2011). Design of A Prompt Gamma Neutron Activation Analysis (PGNAA) System for Boron Neutron Capture Therapy (BNCT) using Monte Carlo Code. Pavia: Disertasi, University of Pavia.
Warfi, R. (2015). Optimasi Kolimator Kolom Termal untuk Fasilitas Uji In Vitro dan In Vivo BNCT di Reaktor Kartini Menggunakan Simulator MCNP-X. Yogyakarta: Universitas Gadjah Mada.
Published
2016-02-28
How to Cite
Dwiputra, M., Harto, A., Sardjono, Y., & Wijaya, G. (2016). Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator. Indonesian Journal of Physics and Nuclear Applications, 1(1), 44-53. https://doi.org/10.24246/ijpna.v1i1.44-53
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This work is licensed under a Creative Commons Attribution-NonCommercial-NoDerivatives 4.0 International License.
Indonesian Journal of Physics and Nuclear Applications is licensed under a Creative Commons Attribution 4.0 International License.
