Study on Concrete Activation Reduction in a PET Cyclotron Vault

Abstract

Background: Concrete activation in cyclotron vaults is a major concern associated with their decommissioning because a considerable amount of activated concrete is generated by secondary neutrons during the operation of cyclotrons. Reducing the amount of activated concrete is important because of the high cost associated with radioactive waste management. This study aims to investigate the capability of the neutron absorbing materials to reduce concrete activation. Materials and Methods: The Particle and Heavy Ion Transport code System (PHITS) code was used to simulate a cyclotron target and room. The dimensions of the room were 457 cm (length), 470 cm (width), and 320 cm (height). Gd2O3, B4C, polyethylene (PE), and borated (5 wt% nat B) PE with thicknesses of 5, 10, and 15 cm and their different combinations were selected as neutron absorbing materials. They were placed on the concrete walls to determine their effects on thermal neutrons. Thin B4C and Gd2O3 were placed between the concrete wall and additional PE shield separately to decrease the required thickness of the additional shield, and the thermal neutron flux at certain depths inside the concrete was calculated for each condition. Subsequently, the optimum combination was determined with respect to radioactive waste reduction, price, and availability, and the total reduced radioactive concrete waste was estimated. Results and Discussion: In the specific conditions considered in this study, the front wall with respect to the proton beam contained radioactive waste with a depth of up to 64 cm without any additional shield. A single layer of additional shield was inefficient because a thick shield was required. Two-layer combinations comprising 0.1-or 0.4-cm-thick B4C or Gd2O3 behind 10 cmthick PE were studied to verify whether the appropriate thickness of the additional shield could be maintained. The number of transmitted thermal neutrons reduced to 30% in case of 0.1 cmthick Gd2O3+10 cm-thick PE or 0.1 cm-thick B4C+10 cm-thick PE. Thus, the thickness of the radioactive waste in the front wall was reduced from 64 to 48 cm. Conclusion: Based on price and availability, the combination of the 10 cm-thick PE+0.1 cmthick B4C was reasonable and could effectively reduce the number of thermal neutrons. The amount of radioactive concrete waste was reduced by factor of two when considering whole concrete walls of the PET cyclotron vault.

Decommissioning of a Medical Cyclotron Vault: The Case Study of the National Cancer Institute of Milano

“…In-situ gamma spectrometry Gamma spectra measured during the in-situ characterization demonstrated the presence of the radionuclides that are reported in Table 1. The list agrees with gamma emitters reported in literature (Masumoto et al 2003;Calandrino et al 2006;Yamaguchi et al 2011;Fujibuchi et al 2015;Vichi et al 2019Vichi et al , 2020Bakhtiari et al 2020).…”