ANALISIS VARIASI MATERIAL PERISAI RADIASI UNTUK FASILITAS BNCT (BORON NEUTRON CAPTURE THERAPY) MENGGUNAKAN SIMULASI MCNP-6

Miqdad Zaki Haidar, NIM.: 18106020040 (2025) ANALISIS VARIASI MATERIAL PERISAI RADIASI UNTUK FASILITAS BNCT (BORON NEUTRON CAPTURE THERAPY) MENGGUNAKAN SIMULASI MCNP-6. Skripsi thesis, UIN SUNAN KALIJAGA YOGYAKARTA.

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Abstract

A research has been done with aims to measure the flux and radiation dose rate over some shielding materials, also to analyze the most effective shielding material. This research is important for radiation workers safety in BNCT facility. The research used MCNP6 (Monte Carlo N-Particle 6) simulator to model the BNCT neutron source and the shielding. The shielding should reduce radiation to less than the dose limit of 10 μSv/hour, which is assumed to be the most conservative limit, i.e. while duration of workers is 2000 hours. Materials used in this research is paraffin, concrete, and BPE (Borated Polyethylene). The research shows the flux value of concrete material 4.98242E+4 particle/cm2.s left side, 8.49752E+4 particle/cm2.s front side, and 6.11110E+4 partikcle/cm2.s right side, with dose rate value of 8974 μSv/Hour left side, 15300 μSv/Hour front side, and 11010 μSv/Hour right side. Flux value BPE 2.27417E+4 particle/cm2.s left side, 7.02884E+4 particle/cm2.s front side, and 2.26976E+4 particle/cm2.s right side, with dose rate value of 4090 μSv/Hour left side, 12642 μSv/Jam front side, and 4084 μSv/Jam right side. Flux value of parafin 5.1732806E+4 particle/cm2.s left side, 6.66068E+4 particle/cm2.s front side, and 5.25806E+4 particle/cm2.s right side, with dose rate value of 9320 μSv/Hour left side, 11996 μSv/Hour front side, and 9472 μSv/Hour right side. Based from data acquired, BPE have better effectivity than the other materials because it has lower dose rate value averaging at 8684 μSv/Hour, meanwhile concrete and paraffin averaging at 11761 μSv/Hour and 10262 μSv/Hour.

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