Boron Neutron Capture Therapy Field Research Articles

New technologies for beam spectrometry, quality assurance, real-time monitoring and microdosimetry in BNCT

Boron neutron capture therapy (BNCT) is a next-generation radiotherapy that combines neutron beams with boron compounds that selectively accumulate in cancer cells. Because this therapy requires neutrons for irradiation, compact accelerator-based neutron source devices that can be installed in hospitals are being developed around the world. Clinical trials of several devices have been conducted in Japan, China, and South Korea. The most advanced device was approved by the Japanese regulatory authorities in 2020. As a result, BNCT with the device is being implemented as an insured therapy for recurrent head and neck cancer at two hospitals in Japan. Thus, the development of accelerator-based neutron generators is proceeding in the field of BNCT. However, much remains to be done to establish this therapy as a common cancer treatment and to make it available worldwide. Dosimetry methods in BNCT are one of the issues. In the treatment, the dose given to a patient by neutron irradiation has to be estimated accurately. However, it is difficult to measure neutrons accurately and in real time. Currently, the neutron dose delivered to patients during treatment is not measured. Instead, the current of the charged particles that irradiate the neutron target material is measured to indirectly evaluate the neutron fluence and dose. In addition, accurate dose estimation based on reactions with neutrons and various elements in a human body uses the Monte Carlo method, which is computationally time-consuming. To address these dosimetry issues, several neutron fluence and dose measurement methods are being developed. This review briefly describes the current dosimetry methods and then presents several methods under development that allow real-time neutron measurements applicable to BNCT.

Boron-Containing Mesoporous Silica Nanoparticles with Effective Delivery and Targeting of Liver Cancer Cells for Boron Neutron Capture Therapy.

Nanocarriers have been researched comprehensively for the development of novel boron-containing agents in boron neutron capture therapy (BNCT). We designed and synthesized a multifunctional mesoporous silica nanoparticle (MSN)-based boron-containing agent. The latter was coated with a lipid bilayer (LB) and decorated with SP94 peptide (SFSIIHTPILPL) on the surface as SP94-LB@BA-MSN. The latter incorporated boric acid (BA) into hydrophobic mesopores, coated with an LB, and modified with SP94 peptide on the LB. SP94-LB@BA-MSN enhanced nano interface tumor-targeting ability but also prevented the premature release of drugs, which is crucial for BNCT because adequate boron content in tumor sites is required. SP94-LB@BA-MSN showed excellent efficacy in the BNCT treatment of HepG-2 cells. In animal studies with tumor-bearing mice, SP94-LB@BA-MSN exhibited a satisfactory accumulation at the tumor site. The boron content reached 40.18 ± 5.41 ppm in the tumor site 4 h after injection, which was 8.12 and 15.51 times higher than those in mice treated with boronated phenylalanine and those treated with BA. For boron, the tumor-to-normal tissue ratio was 4.41 ± 1.13 and the tumor-to-blood ratio was 5.92 ± 0.45. These results indicated that nanoparticles delivered boron to the tumor site effectively while minimizing accumulation in normal tissues. In conclusion, this composite (SP94-LB@BA-MSN) shows great promise as a boron-containing delivery agent for the treatment of hepatocellular carcinoma using BNCT. These findings highlight the potential of MSNs in the field of BNCT.

A novel method for simultaneously measuring boronophenylalanine uptake in brain tumor cells and number of cells using inductively coupled plasma atomic emission spectroscopy

Boron neutron capture therapy (BNCT) combines neutron irradiation with boron compounds that are selectively uptaken by tumor cells. Boronophenylalanine (BPA) is a boron compound used to treat malignant brain tumors. The determination of boron concentration in cells is of great relevance to the field of BNCT. This study was designed to develop a novel method for simultaneously measuring the uptake of BPA by U87 and U251 cells (two brain tumor cell lines) and number of cells using inductively coupled plasma atomic emission spectroscopy (ICP-AES). The results revealed a linear correlation between phosphorus intensity and the numbers of U87 and U251 cells, with correlation coefficients (R2) of 0.9995 and 0.9994, respectively. High accuracy and reliability of phosphorus concentration standard curve were also found. Using this new method, we found that BPA had no significant effect on phosphorus concentration in either U87 or U251 cells. However, BPA increased the boron concentration in U87 and U251 cells in a concentration-dependent manner, with the boron concentration in U87 cells being higher than that in U251 cells. In both U87 and U251 cells, boron was mainly distributed in the cytoplasm and nucleus, accounting for 85% and 13% of the total boron uptake by U87 cells and 86% and 11% of the total boron uptake by U251 cells, respectively. In the U87 and U251 cell-derived xenograft (CDX) animal model, tumor exhibited higher boron concentration values than blood, heart, liver, lung, and brain, with a tumor/blood ratio of 2.87 for U87 cells and 3.11 for U251 cells, respectively. These results suggest that the phosphorus concentration in U87 and U251 cells can represent the number of cells and BPA is easily uptaken by tumor cells as well as in tumor tissue.

Boronotyrosine, a Borylated Amino Acid Mimetic with Enhanced Solubility, Tumor Boron Delivery, and Retention for the Re-emerging Boron Neutron Capture Therapy Field.

Boron neutron capture therapy (BNCT) is a re-emerging binary cellular level cancer intervention that occurs through the interaction of a cancer-specific 10boron (10B) drug and neutrons. We created a new 10B drug, 3-borono-l-tyrosine (BTS), that improves on the characteristics of the main historical BNCT drug 4-borono-l-phenylalanine (BPA). BTS has up to 4 times greater uptake in vitro than BPA and increased cellular retention. Like BPA, BTS uptake is mediated by the l-type amino acid transporter-1 (LAT1) but is less sensitive to natural amino acid competition. BTS can be formulated and bolus dosed at much higher levels than BPA, resulting in 2-3 times greater boron delivery in vivo. Fast blood clearance and greater tumor boron delivery result in superior tumor-to-blood ratios. BTS boron delivery appears to correlate with LAT1 expression. BTS is a promising boron delivery drug that has the potential to improve modern BNCT interventions.

Feasibility study of optical imaging of the boron-dose distribution by a liquid scintillator in a clinical boron neutron capture therapy field.

Evaluation of the boron dose is essential for boron neutron capture therapy (BNCT). Nevertheless, a direct evaluation method for the boron-dose distribution has not yet been established in the clinical BNCT field. To date, even in quality assurance (QA) measurements, the boron dose has been indirectly evaluated from the thermal neutron flux measured using the activation method with gold foil or wire and an assumed boron concentration in the QA procedure. Recently, we successfully conducted optical imaging of the boron-dose distribution using a cooled charge-coupled device (CCD) camera and a boron-added liquid scintillator at the E-3 port facility of the Kyoto University Research Reactor (KUR), which supplies an almost pure thermal neutron beam with very low gamma-ray contamination. However, in a clinical accelerator-based BNCT facility, there is a concern that the boron-dose distribution may not be accurately extracted because the unwanted luminescence intensity, which is irrelevant to the boron dose is expected to increase owing to the contamination of fast neutrons and gamma rays. The purpose of this research was to study the validity of a newly proposed method using a boron-added liquid scintillator and a cooled CCD camera to directly observe the boron-dose distribution in a clinical accelerator-based BNCT field. A liquid scintillator phantom with 10 B was prepared by filling a small quartz glass container with a commercial liquid scintillator and boron-containing material (trimethyl borate); its natural boron concentration was 1wt%. Luminescence images of the boron-neutron capture reaction were obtained in a water tank at several different depths using a CCD camera. The contribution of background luminescence, mainly due to gamma rays, was removed by subtracting the luminescence images obtained using another sole liquid scintillator phantom (natural boron concentration of 0wt%) at each corresponding depth, and a depth profile of the boron dose with several discrete points was obtained. The obtained depth profile was compared with that of calculated boron dose, and those of thermal neutron flux which were experimentally measured or calculated using a Monte Carlo code. The depth profile evaluated from the subtracted images indicated reasonable agreement with the calculated boron-dose profile and thermal neutron flux profiles, except for the shallow region. This discrepancy is thought to be due to the contribution of light reflected from the tank wall. The simulation results also demonstrated that the thermal neutron flux would be severely perturbed by the 10 B-containing phantom if a relatively larger container was used to evaluate a wide range of boron-dose distributions in a single shot. This indicates a trade-off between the luminescence intensity of the 10 B-added phantom and its perturbation effect on the thermal neutron flux. Although a partial discrepancy was observed, the validity of the newly proposed boron-dose evaluation method using liquid-scintillator phantoms with and without 10 B was experimentally confirmed in the neutron field of an accelerator-based clinical BNCT facility. However, this study has some limitations, including the trade-off problem stated above. Therefore, further studies are required to address these limitations.

Characteristics of optically stimulated luminescent dosimeter of beryllium oxide in BNCT irradiation field

The irradiation fields in boron neutron capture therapy (BNCT) consist of neutrons with a wide energy range and undesired high-intensity γ-rays. As the thermoluminescent dosimeter (TLD) used for BNCT dosimetry is out of production, a new type of γ-ray dosimeter is required for the quality assurance (QA) of the treatment. Therefore, in this study, an optically stimulated luminescent dosimeter (OSLD) made of BeO ceramic was selected and its use in the field of BNCT was investigated. The performance of the dosimeter was tested and the effect of a thermal neutron on the γ-ray dose obtained by the OSLD was measured. The OSLDs had a good γ-ray dose linearity of approximately 300 mGy corresponding to the dose in the quality assurance procedure of BNCT, and the thermal neutron sensitivity of the OSLD was less than that of TLD.

Air ionization chamber combined with LiCaAlF6 scintillator for [formula omitted]-ray dose evaluation in boron neutron capture therapy

The boron neutron capture therapy (BNCT) irradiation field includes neutrons with a wide energy range and undesirable high-intensity γ-rays. Although several γ-ray dose evaluation methods have been proposed, eliminating thermal neutron sensitivity is challenging; therefore, we focused on ionization chambers used in other radiotherapy centers. Since the γ-ray dose evaluation with gas-flowing ionization chambers cannot be used in hospitals because of the requirement for treating activated gas and the setting complexity, a method with no gas-flowing ionization chamber is required. In this study, we developed a γ-ray dose evaluation method and combined the air ionization chamber with a LiCAF scintillator. The LiCAF scintillator subtracts the neutron response of the chamber from the total response. In this study, irradiation tests at the BNCT field were conducted to obtain the conversion factor from the events of the LiCAF scintillator owing to the neutron response of the chamber via the conventional γ-ray dose evaluation in the gas flow chambers. In addition, since the conversion factor varies for each irradiation field and cannot be evaluated experimentally in hospitals, we developed a method to calculate the factor by Monte Carlo simulation. Moreover, the sensitivity of the air ionization chamber as neutron energy range and the perturbation for the combined chamber and LiCAF scintillator were discussed.

Status and outlook: Research and development on the neutron source for BNCT

<p indent=0mm>Boron neutron capture therapy (BNCT) is regarded as a revolutionary means for high-accuracy cancer therapy with cell level selectivity. It has unique therapeutic effects for some malignance, such as glioblastoma multiforme, melanoma, recurrent head and neck malignance. It can also be used for cancer treatment of deep organ, such as liver and lung. Even though its principle was proposed about <sc>80 years</sc> ago, it has never been utilized in routine clinic therapy in hospital until 2020. BNCT, as a binary therapeutic method, greatly depends on both high-quality neutron beam and highly target-selective boron drug. Tumor killing in cell-level accuracy can be achieved only when the two elements work together and collaborate closely. To realize such therapeutic effect, BNCT sets high requirement on both neutron source and boron drug. The research on boron drug is related with multi-disciplines, such as chemistry, biology, medicine, radiology, pharmacy and physics. This article focuses on the neutron source for BNCT. Commonly, there are two types of BNCT neutron sources: reactor-based neutron source and accelerator-based neutron source. Up to now, almost all clinic trails of BNCT therapy were performed with the former. Due to limited resources of reactor neutron source, only less than 2000 cases BNCT treatments have been carried out in the world since BNCT method was invented. Thanks for the intense beam proton accelerator development, accelerator based neutron source can provide better beam quality, and especially importantly it can be installed in a hospital environment, which is essential for wide-range application of BNCT. In accelerator-based BNCT field, China has a sound base owing to more than <sc>20 years</sc> research and development in high intensity neutron source for spallation neutron source and accelerator-driven subcritical system. It is meaningful work to transfer the related technology to accelerator-based BNCT (AB-BNCT). In this paper we will firstly introduce the demand on neutron beam specification recommended by IAEA. Then the difficulties in BNCT slow development in passing decades are reviewed. New era of BNCT is coming and one can expect a prosperous future, owing to accelerator-based BNCT. The core technology of AB-BNCT is explored and recent research achievements on BNCT research and development in China are reported. Various types of accelerators now can be used for BNCT facility, including electro-static high voltage, cyclotron and RF linac. Their pro and con are reviewed. Majorly two kinds of neutron generation targets are utilized for BNCT, and the characteristics comparison of these two are analyzed. On the newly constructed BNCT research platform in China, new boron drug research and development work are conducted in many institutes, universities and pharmacy companies. And in recent, a new AB-BNCT facility is under construction for clinic trails in a hospital. So we can have an optimistic viewpoint for BNCT future in China.

IMPROVEMENT OF GAMMA-RAY SUBTRACTION PROCEDURE FOR A CURRENT-MODE NEUTRON DETECTOR WITH A PAIR OF 6Li- AND 7Li-GLASS SCINTILLATORS.

A current-mode neutron detector with a pair of 6Li- and 7Li-glass scintillators has been developed to measure high-flux neutrons in a boron neutron capture therapy field. Neutrons are basically measured by subtracting gamma-ray component using current outputs from the 7Li-glass scintillator. In the present study, the difference in the gamma-ray sensitivity between the 6Li- and 7Li-glass scintillators and the neutron sensitivity for the 7Li-glass scintillator due to the 6Li contamination were also considered to improve the gamma-ray subtraction precision. The gamma-ray subtraction procedure was experimentally investigated in thermal neutron fields with 252Cf and 241Am-Be neutron sources, which have different gamma-ray intensities per unit neutron fluence. A linear relation between neutron fluence and current output was obtained for the neutron detector in the two types of thermal neutron fields with different gamma-ray intensities. It was found that the gamma-ray subtraction procedure is useful for current-mode neutron detectors.

Development of epi-thermal neutron beam intensity detector with 71Ga(n,γ)72Ga reaction for boron neutron capture therapy

Based on the activation method using 71Ga(n,γ)72Ga reaction, a cubic neutron flux intensity detector for epi-thermal neutrons was designed for boron neutron capture therapy (BNCT), and experimentally tested with a prototype detector in a neutron field produced at OKTAVIAN facility of Osaka University, Japan. The experimental test results and related analysis indicated that the performance of the detector was confirmed to be acceptable in the neutron field of BNCT. Practically, the neutron flux intensity mainly covering from 0.5 eV to 10 keV can be measured within 3% by the present detector.

DEPTH DISTRIBUTIONS OF RBE-WEIGHTED DOSE and PHOTON-ISOEFFECTIVE DOSE FOR BORON NEUTRON CAPTURE THERAPY.

Two types of dose were proposed for use in the treatment planning of boron neutron capture therapy (BNCT) for expressing its high relative biological effectiveness (RBE). On one hand, the RBE-weighted dose is the sum of the absorbed doses weighted by fixed RBE for each dose component of BNCT. On the other hand, photon-isoeffective dose is the photon dose to give the same biological effect calculated considering the dose dependence of RBE and the synergetic effect between different types of radiation. In this study, the depth distributions of the two types of dose in a phantom placed at an accelerator-based BNCT field were calculated using Particle and Heavy Ion Transport code System, PHITS, coupled with an extended stochastic microdosimetric kinetic model. Compared with the corresponding RBE-weighted dose, the calculated photon-isoeffective dose was larger at lower absorbed dose and was smaller at higher absorbed dose, primarily due to the consideration of the dose dependence of RBE. In addition, our calculation revealed that the large variance of the intercellular 10B concentration greatly reduces the photon-isoeffective doses. These results suggest that the considerations of the dose dependence of RBE as well as the intercellular heterogeneity in 10B distribution are indispensable for the precise estimate of the biological effect of BNCT.

Abstract ID: 28 Evaluation of silicon and diamond based microdosimetry for boron neutron capture therapy quality assurance

The shift from reactor to accelerator based neutron production has created a renewed interested in Boron Neutron Capture Therapy (BNCT). BNCT is reliant upon the favourable uptake of boron 10 by tumour cells along with the interaction with neutrons to produce high LET fragments (He and Li nuclei) that deposit energy locally within the tumour site. As with any radiation based treatment, Quality Assurance (QA) is crucial. This study extends previous work regarding the application of solid state microdosimetry in the field of BNCT by means of a dedicated Monte Carlo simulation [1] . Geant4 was used to model and optimise the design of silicon on insulator and diamond based microdosimeters [2] , [3] . Detector optimisation in this context pertains to the geometry and materials (i.e., sensitive volume size and probability of neutron activation) to be used in the fabrication of detectors. The study has shown conclusively that whilst the materials currently used in the fabrication of silicon and diamond based microdosimeters are appropriate, there are changes with respect to the sensitive volume thickness that must be addressed. Lastly, the applicability of previously determined correction factors [4] to match the energy deposition response of charged particles within silicon/diamond to water was evaluated within the context of BNCT. Full results and analysis will be presented.

Triple ionization chamber method for clinical dose monitoring with a Be-covered Li BNCT field.

Fast neutron, gamma-ray, and boron doses have different relative biological effectiveness (RBE). In boron neutron capture therapy (BNCT), the clinical dose is the total of these dose components multiplied by their RBE. Clinical dose monitoring is necessary for quality assurance of the irradiation profile; therefore, the fast neutron, gamma-ray, and boron doses should be separately monitored. To estimate these doses separately, and to monitor the boron dose without monitoring the thermal neutron fluence, the authors propose a triple ionization chamber method using graphite-walled carbon dioxide gas (C-CO2), tissue-equivalent plastic-walled tissue-equivalent gas (TE-TE), and boron-loaded tissue-equivalent plastic-walled tissue-equivalent gas [TE(B)-TE] chambers. To use this method for dose monitoring for a neutron and gamma-ray field moderated by D2O from a Be-covered Li target (Be-covered Li BNCT field), the relative sensitivities of these ionization chambers are required. The relative sensitivities of the TE-TE, C-CO2, and TE(B)-TE chambers to fast neutron, gamma-ray, and boron doses are calculated with the particle and heavy-ion transport code system (PHITS). The relative sensitivity of the TE(B)-TE chamber is calculated with the same method as for the TE-TE and C-CO2 chambers in the paired chamber method. In the Be-covered Li BNCT field, the relative sensitivities of the ionization chambers to fast neutron, gamma-ray, and boron doses are calculated from the kerma ratios, mass attenuation coefficient tissue-to-wall ratios, and W-values. The Be-covered Li BNCT field consists of neutrons and gamma-rays which are emitted from a Be-covered Li target, and this resultant field is simulated by using PHITS with the cross section library of ENDF-VII. The kerma ratios and mass attenuation coefficient tissue-to-wall ratios are determined from the energy spectra of neutrons and gamma-rays in the Be-covered Li BNCT field. The W-value is calculated from recoil charged particle spectra by the collision of neutrons and gamma-rays with the wall and gas materials of the ionization chambers in the gas cavities of TE-TE, C-CO2, and TE(B)-TE chambers (10B concentrations of 10, 50, and 100 ppm in the TE-wall). The calculated relative sensitivity of the C-CO2 chamber to the fast neutron dose in the Be-covered Li BNCT field is 0.029, and those of the TE-TE and TE(B)-TE chambers are both equal to 0.965. The relative sensitivities of the C-CO2, TE-TE, and TE(B)-TE chambers to the gamma-ray dose in the Be-covered Li BNCT field are all 1 within the 1% calculation uncertainty. The relative sensitivities of TE(B)-TE to boron dose with concentrations of 10, 50, and 100 ppm 10B are calculated to be 0.865 times the ratio of the in-tumor to in-chamber wall boron concentration. The fast neutron, gamma-ray, and boron doses of a tumor in-air can be separately monitored by the triple ionization chamber method in the Be-covered Li BNCT field. The results show that these doses can be easily converted to the clinical dose with the depth correction factor in the body and the RBE.

Development and characteristics of the HANARO neutron irradiation facility for applications in the boron neutron capture therapy field

The HANARO neutron irradiation facility for various applications in the boron neutron capture therapy (BNCT) field was developed, and its characteristics were investigated. In order to obtain the sufficient thermal neutron flux with a low level of contamination by fast neutrons and gamma rays, a radiation filtering method was adopted. The radiation filter was designed by using a silicon single crystal, cooled by liquid nitrogen, and a bismuth crystal. The installation of the main components of the irradiation facility and the irradiation room was finished. Neutron beam characteristics were measured by using bare and cadmium-covered gold foils and wires. The in-phantom neutron flux distribution was measured for flux mapping inside the phantom. The gamma-ray dose was determined by using TLD-700 thermoluminescence dosimeters. The thermal and fast neutron fluxes and the gamma-ray dose were calculated by using the MCNP code, and they were compared with experimental data. The thermal neutron flux and Cd ratio available at this facility were confirmed to be 1.49 × 109 n cm−2 s−1 and 152, respectively. The maximum neutron flux inside the phantom was measured to be 2.79 × 109 n cm−2 s−1 at a depth of 3 mm in the phantom. The two-dimensional in-phantom neutron flux distribution was determined, and significant neutron irradiation was observed within 20 mm from the phantom surface. The gamma-ray dose rate for the free beam condition was expected to be about 80 cGy h−1. These experimental results were reasonably well supported by calculation using the facility design code. This HANARO thermal neutron facility can be used not only for clinical trials, but also for various pre-clinical studies in the BNCT field.