Mai Nojiri

Boron neutron capture therapy (BNCT) dose calculation often relies on fixed relative biological effectiveness (RBE) and compound biological effectiveness (CBE) values, despite their dependence on beam quality and tumor biology. We developed a microdosimetry-driven framework that predicts cell survival and RBE for BNCT by coupling PHITS lineal energy (T-SED) calculations with the microdosimetric kinetic (MK) model. MK parameters ( α ₀ , β , r _d , y ₀ ) were derived for BNCT relevant cell lines (U87 glioblastoma, NB1RG skin fibroblasts, SAS human squamous carcinoma, and SCC7 murine squamous carcinoma) using low-LET reference datasets curated in the PIDE database and irradiation conditions reproduced in PHITS. The derived parameters successfully reproduced in vitro survival curves for various charged particles across different energies, and when applied to neutron fields representative of BNCT systems (Kyoto University Reactor thermal neutron beam, cyclotron-based epithermal neutron source using a beryllium target, and linear accelerator system using a lithium target), the framework also reproduced measured in vitro data. Predicted RBE at 10% survival (RBE ₁₀ ) agreed with measurements across cell lines and beam qualities, with only a slight deviation for SCC7 under the CICS spectrum and moderate deviations for SAS due to limited and heterogeneous low-LET datasets in PIDE. This method enables spectrum and cell-line specific estimation of biological effect, supporting replacement of fixed RBE/CBE with spectrum aware quantities to improve BNCT dose prescription and safety. The framework can also guide neutron-beam design by providing preliminary RBE estimates prior to construction of the moderator and beam shaping assembly. Incorporating intracellular boron microdistribution in future work is expected to refine CBE estimates and enhance biological accuracy in BNCT treatment planning. This framework provides a physics-based alternative to fixed RBE/CBE values.

The Boron Neutron Capture Therapy (BNCT) irradiation field comprises high-intensity neutrons and gamma-rays that need to be measured separately because of their different biological effects. To conduct appropriate quality assurance (QA), we should measure the neutrons and gamma-rays. In a previous study, an air-ionization chamber combined with a fiber-type neutron monitor was constructed for QA. However, the fibers attached to the chamber were perturbed by neutron scattering. To eliminate the effect of perturbations, we developed an air ionization chamber into which the fiber could be inserted. Irradiation tests in the BNCT irradiation field showed that the perturbation effect was reduced by less than half, and it was confirmed that the system could measure neutrons and gamma-rays simultaneously for weekly QA. The sensitivity of the chamber to neutron energy was evaluated via simulations, and it was found that corrections for sensitivity to high-energy neutrons are required.

Background: Currently, the metal foil activation method is routinely used to measure the neutron output of an accelerator‐based neutron source designed for clinical Boron neutron capture therapy (BNCT). Although this method is well established and has been primarily utilized since the nuclear reactor BNCT era, the process is labour‐intensive and not well‐suited for a busy hospital environment performing routine patient treatment. A replacement neutron detector system that is simple to use and can measure the neutron output in real‐time is necessary.Purpose: Investigation and implementation of an Eu doped LiCaAlF₆ scintillator detector for use in routine quality assurance tests of an accelerator‐based neutron source designed for clinical BNCT.Methods: The response of the scintillator detector was evaluated using the NeuCure BNCT system installed at the Kansai BNCT Medical Center. The measurement repeatability, neutron fluence linearity, and neutron flux dependency of the detector system were evaluated. The beam central axis and off‐axis thermal neutron distribution inside a water phantom were measured and compared with the Monte Carlo treatment planning system (TPS).Results: The scintillator detector system showed high measurement repeatability with a coefficient of variation of less than 0.4%. The detector system showed linear response up to a proton charge of 3.6 C, and the response was stable between a proton current of 0.1 and 1 mA. Both the central axis and off‐axis thermal neutron flux inside a water phantom matched closely with both the metal foil activation method and the Monte Carlo simulation results. The time it took to perform a routine quality assurance test was drastically reduced from 1.5 h down to a few minutes.Conclusion: Implementation of this detector system in the clinic would significantly reduce the time required for routine QA, acceptance, and commissioning, and be a stepping stone to assist expansion of accelerator‐based BNCT systems worldwide.

Accelerator-based boron neutron capture therapy (AB-BNCT) systems that utilize various neutron production methods have been developed in recent years. As the AB-BNCT system is a novel radiation production system, accurately measuring the neutron ambient dose is crucial for evaluating its performance and ensuring its safety. In this study, we designed a neutron detector system capable of measuring the neutron ambient dose H* in a cyclotron-based epithermal neutron source irradiation field using the particle and heavy ion transport code system 3.34 Monte Carlo simulation code. This neutron detector system comprises slow and fast neutron detectors, which are sensitive to energy ranges of 10⁻⁹–10⁻² MeV and 10⁻²–28 MeV, respectively. The response function of this neutron detector system exhibited satisfactory agreement with the conversion coefficients h*, and the dose measurement accuracy was within 6.5%.

Clinical research in boron neutron capture therapy (BNCT) has been conducted worldwide. Currently, the Monte Carlo (MC) method is the only dose calculation algorithm implemented in the treatment planning system for the clinical treatment of BNCT. We previously developed the MC-RD calculation method, which combines the MC method and the removal-diffusion (RD) equation, for fast dose calculation in BNCT. This study aimed to verify the partial-MC-RD calculation method, which utilizes the MC-RD calculation method for a portion of the entire neutron energy range, in terms of calculation accuracy and time as the dose calculation method. We applied the partial-MC-RD calculation method to calculate the total dose for head phantom, comprising soft tissue, brain tissue, and bone. The calculation time and accuracy were evaluated based on the full-MC method. Our accuracy verifications indicated that the partial-MC-RD calculation was mostly comparable with full-MC calculation in the accuracy. However, the assumptions and approximation used in the RD calculation mainly occurred the discrepancy from the full-MC calculation result. Additionally, the partial-MC-RD calculation reduced the time required to approximately 45% for the irradiation to the top and cheek region of head phantom, compared to the full-MC calculation. In conclusion, the MC-RD calculation method can be the basis of a fast dose calculation method in BNCT.

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.