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All Journal Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Lena Rosmayani
Jenderal Soedirman University
Energy Materials Science & Nanotechnology

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A SIMULATION OF IRRADIATION CALCULATIONS ON LUTETIUM-177 PRODUCTION IN RSG-GAS USING U3SI2-AL AND U9MO-AL FUELS Lena Rosmayani; Anis Rohanda; Raden Farzand Abdullatif
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 25, No 2 (2023): June 2023
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2023.6793

Abstract

This research is a simulation of irradiation calculations on the production of the radioisotope Lutetium-177 (177Lu) in the G.A Siwabessy Reactor (RSG-GAS). This study aims to analyze the comparative calculation of 177Lu activity and its purity. One of the production methods of 177Lu in RSG-GAS is carried out by irradiating Lu2O3 targets. This Lu2O3 target irradiation produced the radioisotope 177Lu along with 177mLu as an impurity. For Medical treatment using radioisotopes, the minimum activity for 177Lu is 20 GBq/mg, and the impurity should not exceed 0.1%. Calculations were carried out with thermal neutron flux input at 15 MWt operational power for the RSG-GAS core with U3Si2-Al fuel (density 2.96 gU/cc and 3.55 gU/cc) and U9Mo-Al fuel (density 3.55 gU/cc). Calculations were carried out by simulating 8 days of irradiation using ORIGEN2.1. The results showed that the 177Lu activity resulting from irradiation of Lu2O3 targets at various CIP positions in the U9Mo-Al reactor core was larger than that of the U3Si2-Al core. Until the 30th day, the 177Lu product resulting from irradiation on the U3Si2-Al and U9Mo-Al cores still meets the minimum value of 20 GBq/mg for treatment needs in nuclear medicine, with the activity value of 177Lu resulting from irradiation on the U3Si2-Al core ranging from 241-403 GBq/mg, while the activity of irradiated 177Lu in the U9Mo-Al core ranges from 335-561 GBq/mg. In addition, until the 30th day of decay, 177Lu has a percentage value of 177mLu irradiated in the U9Mo-Al and U3Si2-Al cores of 0.0346% and 0.0344%, respectively. The results are still below the maximum impurity value of 0.1% and thus safe to use as a therapeutic agent. Keywords: 177Lu, Activity, RSG-GAS, ORIGEN2, Irradiation
KARAKTERISASI RADIOACTIVE FISSION WASTE (RFW) DARI PRODUKSI RADIOISOTOP LUTESIUM-177 MENGGUNAKAN PROGRAM ORIGEN2 Lena Rosmayani; Anis Rohanda; R Farzand Abdullatif
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2022.28.2.6672

Abstract

The types and production of radioactive waste continue to increase with the development of the use of technology so that good management is needed so as not to endanger the community or the environment. Production of radioisotopes for nuclear medicine is one of the sources of radioactive waste produced by nuclear reactors. Several radiosotopes produced at the G.A Siwabessy Multipurpose Reactor (RSG–GAS) core include Molybdenum-99, Iodine-125, Iridium-192 and Lutesium-177 (177Lu). The characterization of the 177Lu production waste was carried out using the ORIGEN2 computer program with the aim of obtaining the characteristics of the RFW waste to obtain an appropriate waste treatment technique. In this study, a characterization of radioactive waste was carried out on the production of the radioisotope 177Lu from Target Lu2O3 at various CIP positions at a power of 15 MWt and 30 MWt with a long irradiation time at each power level for 8 and 12 days. In radioisotope production, the target is irradiated in the RSG-GAS core irradiation facility, the target is placed in a quartz ampoule which is then placed in a capsule in aluminum. Then the aluminum inner capsule is inserted into the outer capsule and placed into the irradiation position. After irradiation targeting and then taking 177Lu radioisotope. Sampling of radioisotopes after irradiation forms several types of waste. One of them is radioactive fission waste (RFW) as a by-product with different properties. Apart from the target, radioactive waste can also be generated from the target capsule. Target irradiation makes the target capsule made of quartz and aluminum also activated and becomes radioactive. Therefore, it is necessary to calculate the activity concentration of the waste produced by the radioisotope 177Lu. The results of the characterization of RFW waste using the ORIGEN2 computer program with flux variations at various CIP positions at a power of 15 MWt and 30 MWt with an irradiation period of 8 and 12 days obtained a total concentration of waste activity from the production of 177Lu radioisotope which has an activity concentration between 1,06x1016 – 1,24x1016 Bq/g. Therefore, the RFW radioactive waste produced by 177Lu is classified as moderate radioactive waste based on the Government Regulation on radioactive waste management and requires careful management to ensure safety.
Co-Authors Anis Rohanda Anis Rohanda R Farzand Abdullatif Raden Farzand Abdullatif