Implementation of optically simulated luminescent dosimeter for quality control of gamma ray dose of an accelerator-based neutron source.

BACKGROUND: Neutron beams utilized for performing BNCT are composed of a mixture of neutrons and gamma rays. Although much of the dose delivered to the cancer cells comes from the high LET particles produced by the boron neutron capture reaction, the dose delivered to the healthy tissues from unwanted gamma rays cannot be ignored. With the increase in the number of accelerators for BNCT, a detector system that is capable of measuring gamma ray dose in a mixed neutron/gamma irradiation field is crucial. Currently, BeO TLDs encased in quartz glass are used to measure gamma ray dose in a BNCT irradiation field. However, this type of TLD is no longer commercially available. A replacement dosimetry system is required to perform the recommended ongoing quality assurance of gamma ray measurement for a clinical BNCT system. PURPOSE: The purpose of this study is to evaluate the characteristics of a BeO OSLD detector system under a mixed neutron and gamma ray irradiation field and to assess the suitability of the system for routine quality assurance measurements of an accelerator-based BNCT facility. METHODS: The myOSLD system by RadPro International GmbH was evaluated using the accelerator-based neutron source designed for clinical BNCT (NeuCure BNCT system). The readout constancy, linearity, dose rate effect, and fading effect of the OSLD were evaluated. Free-in-air and water phantom measurements were performed and compared with the TLD results and Monte Carlo simulation results. The PHITS Monte Carlo code was used for this study. RESULTS: The readout constancy was found to be stable over a month-long period and similar to the TLD results. The OSLD readout signal was found to be linear, with a high coefficient of determination (R2 ≥ 0.999) up to a proton charge of 3.6 C. There was no significant signal fading or dose rate dependency. The central axis depth dose and off-axis dose profile measurements agreed with both the TLD and Monte Carlo simulation results, within one standard deviation. CONCLUSION: The myOSLD system was characterized using an accelerator system designed for clinical BNCT. The experimental measurements confirmed the OSLD achieved similar, if not superior to, the currently utilized dosimetry system for routine QA of an accelerator-based BNCT system. The OSLD system would be a suitable replacement for the current TLD system for performing routine QA of gamma ray dose measurement in a BNCT irradiation field.

Neutron flux evaluation algorithm with a combination of Monte Carlo and removal-diffusion calculation methods for boron neutron capture therapy.

BACKGROUND: In Japan, the clinical treatment of boron neutron capture therapy (BNCT) has been applied to unresectable, locally advanced, and recurrent head and neck carcinomas using an accelerator-based neutron source since June of 2020. Considering the increase in the number of patients receiving BNCT, efficiency of the treatment planning procedure is becoming increasingly important. Therefore, novel and rapid dose calculation algorithms must be developed. We developed a novel algorithm for calculating neutron flux, which comprises of a combination of a Monte Carlo (MC) method and a method based on the removal-diffusion (RD) theory (RD calculation method) for the purpose of dose calculation of BNCT. PURPOSE: We present the details of our novel algorithm and the verification results of the calculation accuracy based on the MC calculation result. METHODS: In this study, the "MC-RD" calculation method was developed, wherein the RD calculation method was used to calculate the thermalization process of neutrons and the MC method was used to calculate the moderation process. The RD parameters were determined by MC calculations in advance. The MC-RD calculation accuracy was verified by comparing the results of the MC-RD and MC calculations with respect to the neutron flux distributions in each of the cubic and head phantoms filled with water. RESULTS: Comparing the MC-RD calculation results with those of MC calculations, it was found that the MC-RD calculation accurately reproduced the thermal neutron flux distribution inside the phantom, with the exception of the region near the surface of the phantom. CONCLUSIONS: The MC-RD calculation method is useful for the evaluation of the neutron flux distribution for the purpose of BNCT dose calculation, except for the region near the surface.

Development and evaluation of dose calculation algorithm with a combination of Monte Carlo and point-kernel methods for boron neutron capture therapy.

We developed a 'hybrid algorithm' that combines the Monte Carlo (MC) and point-kernel methods for fast dose calculation in boron neutron capture therapy. The objectives of this study were to experimentally verify the hybrid algorithm and to verify the calculation accuracy and time of a 'complementary approach' adopting both the hybrid algorithm and the full-energy MC method. In the latter verification, the results were compared with those obtained using the full-energy MC method alone. In the hybrid algorithm, the moderation process of neutrons is simulated using only the MC method, and the thermalization process is modeled as a kernel. The thermal neutron fluxes calculated using only this algorithm were compared with those measured in a cubic phantom. In addition, a complementary approach was used for dose calculation in a geometry simulating the head region, and its computation time and accuracy were verified. The experimental verification indicated that the thermal neutron fluxes calculated using only the hybrid algorithm reproduced the measured values at depths exceeding a few centimeters, whereas they overestimated those at shallower depths. Compared with the calculation using only the full-energy MC method, the complementary approach reduced the computation time by approximately half, maintaining nearly same accuracy. When focusing on the calculation only using the hybrid algorithm only for the boron dose attributed to the reaction of thermal neutrons, the computation time was expected to reduce by 95% compared with the calculation using only the full-energy MC method. In conclusion, modeling the thermalization process as a kernel was effective for reducing the computation time.