Boron neutron capture therapy (BNCT): a unique role in radiotherapy with a view to entering the accelerator-based BNCT era.

Boron neutron capture therapy (BNCT) has a unique property of tumor-cell-selective heavy-particle irradiation. BNCT can form large dose gradients between cancer cells and normal cells, even if the two types of cells are mingled at the tumor margin. This property makes it possible for BNCT to be used for pre-irradiated locally recurrent tumors. Shallow-seated, locally recurrent lesions have been treated with BNCT because of the poor penetration of neutrons in the human body. BNCT has been used in clinical studies for recurrent malignant gliomas and head and neck cancers using neutron beams derived from research reactors, although further investigation is warranted because of the small number of patients. In the latter part of this review, the development of accelerator-based neutron sources is described. BNCT for common cancers will become available at medical institutes that are equipped with an accelerator-based BNCT system. Multiple metastatic lung tumors have been investigated as one of the new treatment candidates because BNCT can deliver curative doses of radiation to the tumors while sparing normal lung tissue. Further basic and clinical studies are needed to move toward an era of accelerator-based BNCT when more patients suffering from refractory cancers will be treated.

Boron neutron capture therapy (BNCT) translational studies in the hamster cheek pouch model of oral cancer at the new "B2" configuration of the RA-6 nuclear reactor.

Boron neutron capture therapy (BNCT) is based on selective accumulation of B-10 carriers in tumor followed by neutron irradiation. We demonstrated, in 2001, the therapeutic effect of BNCT mediated by BPA (boronophenylalanine) in the hamster cheek pouch model of oral cancer, at the RA-6 nuclear reactor. Between 2007 and 2011, the RA-6 was upgraded, leading to an improvement in the performance of the BNCT beam (B2 configuration). Our aim was to evaluate BPA-BNCT radiotoxicity and tumor control in the hamster cheek pouch model of oral cancer at the new "B2" configuration. We also evaluated, for the first time in the oral cancer model, the radioprotective effect of histamine against mucositis in precancerous tissue as the dose-limiting tissue. Cancerized pouches were exposed to: BPA-BNCT; BPA-BNCT + histamine; BO: Beam only; BO + histamine; CONTROL: cancerized, no-treatment. BNCT induced severe mucositis, with an incidence that was slightly higher than in "B1" experiments (86 vs 67%, respectively). BO induced low/moderate mucositis. Histamine slightly reduced the incidence of severe mucositis induced by BPA-BNCT (75 vs 86%) and prevented mucositis altogether in BO animals. Tumor overall response was significantly higher in BNCT (94-96%) than in control (16%) and BO groups (9-38%), and did not differ significantly from the "B1" results (91%). Histamine did not compromise BNCT therapeutic efficacy. BNCT radiotoxicity and therapeutic effect at the B1 and B2 configurations of RA-6 were consistent. Histamine slightly reduced mucositis in precancerous tissue even in this overly aggressive oral cancer model, without compromising tumor control.

Boron neutron capture synovectomy (BNCS) as a potential therapy for rheumatoid arthritis: radiobiological studies at RA-1 Nuclear Reactor in a model of antigen-induced arthritis in rabbits.

Rheumatoid arthritis is a chronic autoimmune pathology characterized by the proliferation and inflammation of the synovium. Boron neutron capture synovectomy (BNCS), a binary treatment modality that combines the preferential incorporation of boron carriers to target tissue and neutron irradiation, was proposed to treat the pathological synovium in arthritis. In a previous biodistribution study, we showed the incorporation of therapeutically useful boron concentrations to the pathological synovium in a model of antigen-induced arthritis (AIA) in rabbits, employing two boron compounds approved for their use in humans, i.e., decahydrodecaborate (GB-10) and boronophenylalanine (BPA). The aim of the present study was to perform low-dose BNCS studies at the RA-1 Nuclear Reactor in the same model. Neutron irradiation was performed post intra-articular administration of BPA or GB-10 to deliver 2.4 or 3.9 Gy, respectively, to synovium (BNCS-AIA). AIA and healthy animals (no AIA) were used as controls. The animals were followed clinically for 2 months. At that time, biochemical, magnetic resonance imaging (MRI) and histological studies were performed. BNCS-AIA animals did not show any toxic effects, swelling or pain on palpation. In BNCS-AIA, the post-treatment levels of TNF-α decreased in four of six rabbits and IFN-γ levels decreased in five of six rabbits. In all cases, MRI images of the knee joint in BNCS-AIA resembled those of no AIA, with no necrosis or periarticular effusion. Synovial membranes of BNCS-AIA were histologically similar to no AIA. BPA-BNCS and GB-10-BNCS, even at low doses, would be therapeutically useful for the local treatment of rheumatoid arthritis.

The hamster cheek pouch model for field cancerization studies.

External carcinogens, such as tobacco and alcohol, induce molecular changes in large areas of oral mucosa, which increase the risk of malignant transformation. This condition, known as 'field cancerization', can be detected in biopsy specimens using histochemical techniques, even before histological alterations are identified. The efficacy of these histochemical techniques as biomarkers of early cancerization must be demonstrated in appropriate models. The hamster cheek pouch oral cancer model, universally employed in biological studies and in studies for the prevention and treatment of oral cancer, is also an excellent model of field cancerization. The carcinogen is applied in solution to the surface of the mucosa and induces alterations that recapitulate the stages of cancerization in human oral mucosa. We have demonstrated that the following can be used for the early detection of cancerized tissue: silver staining of nucleolar organizer regions; the Feulgen reaction to stain DNA followed by ploidy analysis; immunohistochemical analysis of fibroblast growth factor-2, immunohistochemical labeling of proliferating cells to demonstrate an increase of epithelial cell proliferation in the absence of inflammation; and changes in markers of angiogenesis (i.e. those indicating vascular endothelial growth factor activity, endothelial cell proliferation and vascular density). The hamster cheek pouch model of oral cancer was also proposed and validated by our group for boron neutron capture therapy studies for the treatment of oral cancer. Clinical trials of this novel treatment modality have been performed and are underway for certain tumor types and localizations. Having demonstrated the efficacy of boron neutron capture therapy to control tumors in the hamster cheek pouch oral cancer model, we adapted the model for the long-term study of field cancerized tissue. We demonstrated the inhibitory effect of boron neutron capture therapy on tumor development in field cancerized tissue with acceptable levels of mucositis, a dose-limiting side-effect.

Synthesis and in vitro evaluation of thiododecaborated α, α- cycloalkylamino acids for the treatment of malignant brain tumors by boron neutron capture therapy.

Boron-neutron capture therapy (BNCT) is an attractive technique for cancer treatment. As such, α, α-cycloalkyl amino acids containing thiododecaborate ([B12H11](2-)-S-) units were designed and synthesized as novel boron delivery agents for BNCT. In the present study, new thiododecaborate α, α-cycloalkyl amino acids were synthesized, and biological evaluation of the boron compounds as boron carrier for BNCT was carried out.

Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

PURPOSE: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. METHODS: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. RESULTS: Local mixed-field thermal neutron sensitivities and global thermal and mixed-field thermal neutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermal neutron sensitivity showed agreement with pure thermal neutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermal neutron sensitivity of 1.95 ± 0.05 × 10(-21) A n(-1)[middle dot]cm² [middle dot]s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. CONCLUSIONS: The usefulness of the CNEA Rh SPND for the on-line local measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.

A new approach to dose estimation and in-phantom figure of merit measurement in BNCT by using artificial neural networks.

In-phantom figures of merit of the radiobiological dose distribution are the main criteria for evaluation of the boron neutron capture therapy (BNCT) plan and neutron beam evaluation. Since in BNCT there are several reactions, which contribute to the total dose of the tissue, the calculation of the dose distribution is complicated and requires lengthy and time-consuming simulations. Any changes in the beam shaping assembly (BSA) design would lead to the change of the neutron/gamma spectrum at exit of therapeutic window. As a result of any changes in the beam spectrum, the dose distribution in the tissue will be altered; therefore, another set of lengthy and time-consuming simulations to recalculate the dose distribution would have to be performed. This study proposes a method that applies artificial neural network (ANN) for quick dose prediction in order to avoid lengthy calculations. This method allows us to estimate the depth-dose distribution and in-phantom figures of merit for any energy spectrum without performing a complete Monte Carlo code (MCNP) simulation. To train the ANNs for modeling the depth-dose distribution, this study used a database containing 500 simulations of the neutron depth-dose distribution and 280 simulations of the gamma depth-dose distribution. The calculations were carried out by the MCNP for various mono-energetic neutrons, ranging from thermal up to 10 MeV energy and 280 gamma energy group, ranging from 0.01 MeV up to 20 MeV, through the SNYDER head phantom which is located at the exit of the BSA. The trained ANN was capable of establishing a map between the neutron/gamma beam energy and the dose distribution in the phantom as an input and a response, respectively. The current method is founded upon the observation that the dose which is released by the beam of composite energy spectrum can be decomposing into the various energy components which make the neutron/gamma spectrum. Therefore, in this procedure the neutron/gamma energy spectrum was converted into several energy groups and dose response of each group was predicted by the trained ANN. Total dose distribution of the entire spectrum is equal to summation of dose response of each group. If the neutron/gamma spectrum as an input changes, the dose response of that as an output can be predicted by the trained ANN in no time rather than hours or days by MCNP simulations. To check the validity of this method, this study compared full calculation of the depth-dose distribution with prediction of ANN for that. The result of this comparison shows that artificial neural networks model the dose distribution in phantom successfully and result in a great accurate prediction.

A bi-tapered and air-gapped beam shaping assembly used for AB-BNCT.

The 7Li(p,n)7Be reaction, which leads to a soft neutron field, is often chosen as the neutron producing reaction used for accelerator-based boron neutron capture therapy (AB-BNCT). This study aims to design a compact beam shaping assembly (BSA) and auxiliary system for a 7Li(p,n)7Be reaction-based neutron source and to evaluate the relationship between the BSA design and the consequent neutron beam quality for further optimization. In this study, five types of moderator shapes for the BSA model were designed. Both the in-air and in-phantom figures of merit were considered to evaluate the performance of the BSA designs. It was found that the BSA with a bi-tapered and air-gapped design could generate a high-intensity epithermal neutron beam, which could be used to treat deep-seated brain tumors within a reasonable time.

Optimized beam shaping assembly for a 2.1-MeV proton-accelerator-based neutron source for boron neutron capture therapy.

Boron Neutron Capture Therapy (BNCT) is facing a new era where different projects based on accelerators instead of reactors are under development. The new facilities can be placed at hospitals and will increase the number of clinical trials. The therapeutic effect of BNCT can be improved if a optimized epithermal neutron spectrum is obtained, for which the beam shape assembly is a key ingredient. In this paper we propose an optimal beam shaping assembly suited for an affordable low energy accelerator. The beam obtained with the device proposed accomplishes all the IAEA recommendations for proton energies between 2.0 and 2.1 MeV. In addition, there is an overall improvement of the figures of merit with respect to BNCT facilities and previous proposals of new accelerator-based facilities.

Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz.

To establish Boron Neutron Capture Therapy (BNCT) for non-resectable liver metastases and for in vitro experiments at the TRIGA Mark II reactor at the University of Mainz, Germany, it is necessary to have a reliable dose monitoring system. The in vitro experiments are used to determine the relative biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using FLUKA, a multipurpose Monte Carlo transport code. The pin-diode is augmented by a lithium fluoride foil. This foil converts the neutrons into alpha and tritium particles which are products of the (7)Li(n,α)(3)H-reaction. These particles are detected by the diode and their amount correlates to the neutron fluence directly. Results and discussion. Gold foil activation and the pin-diode are reliable fluence measurement systems for the TRIGA reactor, Mainz. Alanine dosimetry of the photon field and charged particle field from secondary reactions can in principle be carried out in combination with MC-calculations for mixed radiation fields and the Hansen & Olsen alanine detector response model. With the acquired data about the background dose and charged particle spectrum, and with the acquired information of the neutron flux, we are capable of calculating the dose to the tissue. Conclusion. Monte Carlo simulation of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also speculate on sensitizing alanine to thermal neutrons by adding boron compounds.

Design and construction of an accelerator-based boron neutron capture therapy (AB-BNCT) facility with multiple treatment rooms at the Southern Tohoku BNCT Research Center.

Installation of an accelerator-based boron neutron capture therapy (AB-BNCT) system was started in April 2014 at the Southern Tohoku BNCT Research Center (STBRC), and clinical trials began in January 2016. There are two treatment rooms, which have same specifications, and the beam quality equivalency was confirmed both rooms. Here, we describe the design and construction of the first hospital-based AB-BNCT facility in the world with multiple treatment rooms.

Simulation for improved collimation system of gamma-ray telescope system for boron neutron capture therapy at Kyoto University Reactor.

A gamma-ray telescope system has been used at Heavy Water Neutron Irradiation Facility at Kyoto University Reactor mainly for boron neutron capture therapy (BNCT) for multiple hepatic tumors. This system has been improved to accommodate BNCT for other sites, such as brain, head and neck, lung, breast, etc. Simulation for the collimation system was performed. It revealed that the effective telescope field-of-view could be expanded from approximately 3-21 cm.

Microdosimetric quantities of an accelerator-based neutron source used for boron neutron capture therapy measured using a gas-filled proportional counter.

Boron neutron capture therapy (BNCT) is an emerging radiation treatment modality, exhibiting the potential to selectively destroy cancer cells. Currently, BNCT is conducted using a nuclear reactor. However, the future trend is to move toward an accelerator-based system for use in hospital environments. A typical BNCT radiation field has several different types of radiation. The beam quality should be quantified to accurately determine the dose to be delivered to the target. This study utilized a tissue equivalent proportional counter (TEPC) to measure microdosimetric and macrodosimetric quantities of an accelerator-based neutron source. The micro- and macro-dosimetric quantities measured with the TEPC were compared with those obtained via the the particle and heavy ion transport code system (PHITS) Monte Carlo simulation. The absorbed dose from events >20 keV/µm measured free in air for a 1-h irradiation was calculated as 1.31 ± 0.02 Gy. The simulated result was 1.41 ± 0.07 Gy. The measured and calculated values exhibit good agreement. The relative biological effectiveness (RBE) that was evaluated from the measured microdosimetric spectrum was calculated as 3.7 ± 0.02, similar to the simulated value of 3.8 ± 0.1. These results showed the PHITS Monte Carlo simulation can simulate both micro- and macro-dosimetric quantities accurately. The RBE was calculated using a single-response function, and the results were compared with those of several other institutes that used a similar method. However, care must be taken when using such a single-response function for clinical application, as it is only valid for low doses. For clinical dose ranges (i.e., high doses), multievent distribution inside the target needs to be considered.

The E LiBANS project: Thermal and epithermal neutron sources based on a medical Linac.

The E_LIBANS project (INFN) aims at producing neutron facilities for interdisciplinary irradiation purposes among which pre-clinical research for BNCT. After the successful setting-up of the thermal neutron source based on a medical LINAC, a similar apparatus for epithermal neutrons has been developed. Both structures are based on an Elekta 18 MV coupled with a photoconverter-moderator system which deploys the (γ,n) reaction to convert the X-rays into neutrons. This communication describes the two neutron sources and the results obtained in their characterization.

Implications of radiation microdosimetry for accelerator-based boron neutron capture therapy: a radiobiological perspective.

Boron neutron capture therapy (BNCT) has great potential to selectively destroy cancer cells while sparing surrounding normal cells. The basic concept of BNCT was developed in the 1930s, but it has not yet been commonly used in clinical practice, even though there is now a large number of experimental and translational studies demonstrating its marked therapeutic potential. With the development of neutron accelerators that can be installed in medical institutions, accelerator-based BNCT is expected to become available at several medical institutes around the world in the near future. In this commentary, from the point of view of radiation microdosimetry, we discuss the biological effects of BNCT, especially the underlying mechanisms of compound biological effectiveness. Radiobiological perspectives provide insight into the effectiveness of BNCT in creating a synergy effect in the field of clinical oncology.

Dependence of neutrons generated by 7Li(p,n) reaction on Li thickness under free-air condition in accelerator-based boron neutron capture therapy system employing solid-state Li target.

PURPOSE: An accelerator-based boron neutron capture therapy (BNCT) system with a solid-state Li target is reported to have degradation of the Li target. The degradation reduces the Li thickness, which may change spectra of the generated neutrons corresponding to the Li thickness. This study aims to examine the relationship between the Li thickness and the generated neutrons and to investigate the effects of the Li thickness on the absorbed dose in BNCT. METHOD: The neutron energy spectra were calculated via Monte Carlo simulation for Li thicknesses ranging from 20 to 150 µm. Using the system, the saturated radioactivity of gold induced by reactions between 197Au and the generated neutrons was evaluated with the simulation and the measurement, and those were compared. Additionally, for each Li thickness, the saturated radioactivity was compared with the number of generated neutrons. The absorbed doses delivered by 10B(n,α)7Li, 14N(n,p)14C, 1H(n, g)2H, and (n,n') reactions in water were also calculated for each Li thickness. RESULTS: The measurement and simulation indicated a reduction in the number of neutrons due to the degradation of the Li target. However, the absorbed doses were comparable for each Li thickness when the requisite number of neutrons for BNCT was delivered. Additionally, the saturated radioactivity of 198Au could be a surrogate for the number of neutrons even if the Li thickness was varied. CONCLUSIONS: No notable effect to the absorbed dose was observed when required neutron fluence was delivered in the BNCT even if the degradation of the Li was observed.

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.

Characterization of the relationship between neutron production and thermal load on a target material in an accelerator-based boron neutron capture therapy system employing a solid-state Li target.

An accelerator-based boron neutron capture therapy (BNCT) system that employs a solid-state Li target can achieve sufficient neutron flux derived from the 7Li(p,n) reaction. However, neutron production is complicated by the large thermal load expected on the target. The relationship between neutron production and thermal load was examined under various conditions. A target structure for neutron production consists of a Li target and a target basement. Four proton beam profiles were examined to vary the local thermal load on the target structure while maintaining a constant total thermal load. The efficiency of neutron production was evaluated with respect to the total number of protons delivered to the target structure. The target structure was also evaluated by observing its surface after certain numbers of protons were delivered. The yield of the sputtering effect was calculated via a Monte Carlo simulation to investigate whether it caused complications in neutron production. The efficiency of neutron production and the amount of damage done depended on the proton profile. A more focused proton profile resulted in greater damage. The efficiency decreased as the total number of protons delivered to the target structure increased, and the rate of decrease depended on the proton profile. The sputtering effect was not sufficiently large to be a main factor in the reduction in neutron production. The proton beam profile on the target structure was found to be important to the stable operation of the system with a solid-state Li target. The main factor in the rate of reduction in neutron production was found to be the local thermal load induced by proton irradiation of the target.

Radiation Chemical Yield of Hydroxyl Radicals for Accelerator-based Boron Neutron Capture Therapy: Dose Assessment of 10B(n,α)7Li Reaction Using Coumarin-3-Carboxilic Solution.

Evaluation of the characteristics of accelerator-based thermal neutron fields is recognized as an important issue when discussing the effectiveness of boron neutron capture therapy (BNCT). In this study, we propose that the radiation chemical yield (G value) of hydroxyl radicals (Goh•) can be considered a universal parameter for the description of the accelerator-based thermal neutron field. The Goh• of the 10B(n,α)7Li reaction was quantitatively evaluated using an aqueous coumarin-3-carboxylic acid (3CCA) solution, and was discriminated from that of contaminations (i.e., γ rays and fast neutrons). The Goh• of the 10B(n,α)7Li reaction was 0.107 ± 0.004 OH•/100 eV, which is almost equivalent to that exposed to α particles with an energy of 6.0 MeV. Since the Goh• of γ rays from a 60Co source is 2.03 ± 0.05 OH•/100 eV, this lower value suggests that indirect action by the 10B(n,α)7Li reaction is not dominant in BNCT. However, our results indicate that one can assess the 60Co equivalent dose of the 10B(n,α)7Li reaction in water from the Goh• derived using aqueous 3CCA solution in the accelerator-based thermal neutron field.

Radiation shielding evaluation of the BNCT treatment room at THOR: a TORT-coupled MCNP Monte Carlo simulation study.

This study investigates the radiation shielding design of the treatment room for boron neutron capture therapy at Tsing Hua Open-pool Reactor using "TORT-coupled MCNP" method. With this method, the computational efficiency is improved significantly by two to three orders of magnitude compared to the analog Monte Carlo MCNP calculation. This makes the calculation feasible using a single CPU in less than 1 day. Further optimization of the photon weight windows leads to additional 50-75% improvement in the overall computational efficiency.