Neutron Capture Therapy Facility Research Articles

- Beam Neutron Optimization for Boron Neutron Capture Therapy (BNCT) facility

- Beam Neutron Optimization for Boron Neutron Capture Therapy (BNCT) facility

NeuTHOR Station-A Novel Integrated Platform for Monitoring BNCT Clinical Treatment, Animal and Cell Irradiation Study at THOR.

(1) Background: A well-established Boron Neutron Capture Therapy (BNCT) facility includes many essential systems, which are the epithermal neutron beam system, on-line monitoring system (OMS), QA/QC (quality assurance or quality control) system, boron concentration (BC) measurement system, and treatment planning system (TPS). Accurate data transmission, monitoring, and deposition among these systems are of vital importance before, during, and after clinical, animal, and cell BNCT irradiation. This work developed a novel integrated platform NeuTHOR Station (NeuTHORS) for BNCT at Tsing Hua Open-pool Reactor (THOR). Apart from the data of the OMS and QA/QC system, the data of BC and TPS can be loaded on NeuTHORS before BNCT clinical, animal, and cell irradiation. (2) Methods: A multi-paradigm computer programming language c# (c sharp) was used to develop the integrated platform NeuTHORS. The design of NeuTHORS is based on the standard procedures of BNCT treatment or experiment at THOR. Moreover, parallel testing with OMS-BNCT (the former OMS) and QA/QC of THOR was also performed for more than 70 times to verify the validation of NeuTHORS. (3) Results: According to the comparisons of the output, NeuTHORS and OMS-BNCT and QA/QC of THOR show very good consistency. NeuTHORS is now installed on an industrial PC (IPC) and successfully performs the monitoring of BNCT Treatment at THOR. Patients' f BC and TPS data are also input into NeuTHORS and stored on IPC through an internal network from BC measurement room and TPS physicist. Therefore, the treatment data of each patient can be instantaneously established after each BNCT treatment for further study on BNCT. NeuTHORS can also be applied on data acquisition for a BNCT-related study, especially for animal or cell irradiation experiments. (4) Conclusions: A novel integrated platform NeuTHOR Station for monitoring BNCT clinical treatment and animal and cell irradiation study has been successfully established at THOR. With this platform, BNCT radiobiology investigations will be efficiently performed and a thorough data storage and analysis system of BNCT treatments or experiments can thus be systematically built up for the further investigation of BNCT at THOR.

Current development status of iBNCT001, demonstrator of a LINAC-based neutron source for BNCT

The iBNCT project aims to develop “iBNCT001,” a demonstration device of the linac-based neutron irradiation facility for boron neutron capture therapy (BNCT) application. iBNCT001 generates an epithermal neutron beam by irradiating 8 MeV protons accelerated by a linac onto a beryllium target. Currently, the linac can drive an average proton current of 2.1 mA. Several experiments were performed using a water phantom to confirm the main physical characteristics of the neutron beam produced at the irradiation position. The measurement results demonstrated that the maximum thermal neutron flux achievable in the phantom volume was approximately 1.36 × 10 9 cm − 2 s − 1 when a normal beam collimator with a 120 mm diameter was used. This neutron beam intensity was sufficient to complete the irradiation within 30 min using the BNCT approach. In addition to normal beam collimators, extended collimators that protrude 100 mm from the wall were developed. By using an extended collimator, it is possible to prevent interference of the patient’s body with the wall when irradiating head and neck cancers. The measurement results for the extended collimator demonstrated that irradiation with the collimator could be completed within 1 h when the neutron beam is generated with an average proton current of 2.1 mA.

Evaluation of depth-dose profiles in a water phantom at the BNCT facility at BINP

In this study, we present depth-dose profiles measured with a water phantom at the boron neutron-capture therapy (BNCT) facility at Budker Institute of Nuclear Physics (BINP). The presented results demonstrate that the proposed design of radiation detector with an optical fiber readout, which includes three different sensors (the first based on a plastic scintillator enriched with boron, the second based on a simple plastic scintillator, and the third having no scintillator at all), enables measurement of neutron flux, as well as estimations of the dose induced by gamma rays. The results of simulations with the GEANT4 package demonstrated good agreement with the experimental data and can be used for further system improvement.

Design of the input power coupler for Boron Neutron Capture Therapy Drift Tube Linac

An accelerator-based Boron Neutron Capture Therapy (BNCT) facility has been built at the campus of China Spallation Neutron Source (CSNS). To upgrade it, a Drift Tube Linac (DTL) is designed to provide a 40 mA, 25% duty factor, 10.1 MeV beam for the target. Two opposing power couplers are employed to deliver forward peak power up to 640 kW for the beam and DTL cavity. A new coaxial power coupler with a single Tristan type window is designed to meet the power and coupling coefficient requirements. In this paper, the detailed RF optimization process of the coupler is described. According to the RF simulation results, the multipacting (MP) phenomenon at critical locations is simulated by the CST PS (Particle Studio). Simulations of thermal and mechanical are also carried out, the cooling design to provide a large margin for RF heating is performed based on the simulation results.

Conceptual Shield Design for Boron Neutron Capture Therapy Facility Using Monte Carlo N-Particle Extended Simulator with Kartini Research Reactor as Neutron Source

The research aims to measure the radiation dose rate over the radiation shielding which is made of paraffin and aluminium and to determine the best shield material for the safety of radiation workers. The examination used MCNP (Monte Carlo N-Particle) simulator to model the BNCT neutron source and the shield. The shield should reduce radiation to less than the dose limit of 10.42 µSv/h, which is assumed to be the most conservative limit when the duration of workers is 1920 h. The first design resulted in a radiation dose rate which was still greater than the limit. Therefore, optimization was done by adding the lead on the outer part of the shield. After optimization by adding the lead with certain layers, the radiation dose rate decreased, with the largest dose being 57.60 µSv/h. Some locations over the limit could be overcome by other radiation protection aspects such as distance and time. The paraffin blocks were covered by aluminium to keep the shield structure. The lead was used to absorb the gamma ray which resulted from the interaction between the neutrons and aluminium.

Comparison of different methods for the shielding analysis of an AB-BNCT facility based on the Be(p,xn) reaction with 30-MeV protons

Five shielding calculation methods were employed to estimate the dose rate distribution around an accelerator-based boron neutron capture therapy facility. Their performances were compared in terms of accuracy and efficiency. The results indicate that the hybrid deterministic/Monte Carlo method is the most efficient in the context of accurate modeling and simulation, whereas the analytical approximation with pre-generated source terms and attenuation lengths is preferable in the design phase because of its simplicity and ease of verification while retaining a reasonable accuracy.

Neutron yields of Be-9(p,xn) reactions and beam characterization for accelerator-based boron neutron capture therapy facility using MCNP6, PHITS, and GEANT4 simulation results

Neutron yields from 9Be(p,xn) reactions with a 2-cm thick beryllium target bombarded by proton beams of 5, 6, 7, 8, 9, and 10 MeV were investigated. Neutron beam characteristics through a pilot design of beam shaping assembly were calculated by using MCNP6, PHITS, and GEANT4. The neutron yield with a 10 MeV beam by MCNP6 was 7.7 times larger than values by the other codes, but close to the experimental data. The neutron spectrum and directionality were different among the codes. These differences could stem from different cross-section libraries and models used in the codes. The epithermal flux at the beam irradiation port by MCNP6 was 1.4 × 109 cm−2 s−1 for a proton current of 8 mA hitting the target. The ratio of thermal and epithermal flux was less than 0.01 and the forward directionality was 0.78. However, we suggest a new criterion regarding a ratio of fast and epithermal flux.

A design on the collimator for boron neutron capture therapy (BNCT) research facility at the thermal column of TRIGA MARK II

The development of the Boron Neutron Capture Therapy (BNCT) facility in Malaysia can be performed at the thermal column of the Malaysia research reactor. TRIGA MARK II is one of the facilities that can provide neutron source for BNCT facility. The specification of neutron flux and the gamma dose rate must consider for the development of the BNCT facility as a safety precaution for this research. Based on previous research, the thermal column identified as a suitable place for BNCT facility. To design the neutron collimator for BNCT purpose, the characterization of material towards thermal neutron flux explored using TLD and Microspec-6 and the collimator design was simulate using Monte Carlo N-Particle (MCNP) software based on the characterize materials in order to produce high thermal neutron flux. The combination of lead, HDPE, 30% borated polyethylene and aluminium as collimator design D1 simulate the highest thermal neutron 1.5770 x 109 neutron.cm−2s−1 and suitable for BNCT research purpose at the thermal column.

Optical diagnostics of negative ion beam transport.

A diagnostic complex for monitoring the position, propagation direction, and angular distribution dispersion of a particle beam planned for application in the Boron Neutron Capture Therapy facility is described in this paper. For the beam position and direction, the precision is, respectively, 0.1 mm and 1 mrad at 10 mA CW H- beam with energy of about 35 keV and a diameter of the order of 10 mm. The energy spread and angular divergence were measured within the accuracy of about 100 eV and 3 mrad, respectively. The acceptable precision of about 1 mm for the beam position is obtained at a relatively short exposure to 10 ms. To increase the radiation intensity of the beam, the addition of various gases was studied. The addition of gas decreases the beam width and increases the negative ions stripping.

A multichannel neutron flux monitoring system for a boron neutron capture therapy facility

The progress in the design of new accelerator-based neutron sources for boron neutron capture therapy (BNCT) generated new requirements to the detector systems to monitor the particle flux during the irradiation of patients. In the current research, we evaluated basic parameters of a twin scintillator-over-fiber system with a silicon photo-multiplier readout based on boron-enriched and boron-free plastic scintillators. The presented results showed that the proposed system could be used for monitoring of neutron flux, and the designed detector system demonstrated good linearity, up to ∼ 2 mA, limited by accelerator specifications at that time.

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.

MONTE CARLO N PARTICLE EXTENDED (MCNPX) RADIATION SHIELD MODELLING ON BORON NEUTRON CAPTURE THERAPY FACILITY USING D-D NEUTRON GENERATOR 2.4 MeV

Based Studies were carried out to analyze the internal dose of radiation for workers at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and a room that was actually designed before. This internal dose analyzation included interaction between neutrons and air. The air contained N2 (72%), O2 (20%), Ar (0.93%), CO2, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be below the dose limit for radiation workers which is an amount of 20 mSv/years. From the particles that are present in the air, only Nitrogen and Argon can change into radioactive element. Nitrogen-14 activated to Carbon-14, Nitrogen-15 activated to Nitrogen-16, and Argon-40 activated to Argon-41. Calculation using tally facility in Monte Carlo N Particle version Extended (MCNPX) program for calculated Neutron flux in the air 3.16x107 Neutron/cm2s. The room design in the cancer facility has a measurement of 200 cm in length, 200 cm in width, and 166.40 cm in height. Neutron flux can be used to calculate the reaction rate which is 80.1x10-2 reaction/cm3s for carbon-14 and 8.75x10-5 reaction/cm3s. The internal dose exposed to the radiation worker is 9.08E-9 µSv.

Beam Shaping Assembly Optimization for Boron Neutron Capture Therapy Facility Based on Cyclotron 30 MeV as Neutron Source

A design of beam shaping assembly (BSA) installed on cyclotron 30 MeV model neutron source for boron neutron capture therapy (BNCT) has been optimized using simulator software of Monte Carlo N-Particle Extended (MCNPX). The Beryllium target with thickness of 0.55 cm is simulated to be bombarded with 30 MeV of proton beam. In this design, the parameter regarding beam characteristics for BNCT treatment has been improved, which is ratio of fast neutron dose and epithermal neutron flux. TiF3 is replaced to 30 cm of 27Al as moderator, and 1.5 cm of 32S is combined with 28 cm of 60Ni as neutron filter. Eventually, this design produces epithermal neutron flux of 2.33 × 109, ratio between fast neutron dose and epithermal neutron flux of 2.12 × 10-13,ratio between gamma dose and epithermal neutron flux of 1.00 × 10-13, ratio between thermal neutron flux and epithermal neutron flux is 0.047, and ration between particle current and total neutron flux is 0.56.

Accelerator-based neutron source for boron neutron capture therapy

A beam shaping assembly (BSA), a neutron moderator system, is a key component of an accelerator based boron neutron capture therapy (BNCT) facility. The neutron energy recommended in the IAEA TECDOC-1223 for the BNCT is much lower than the energy of the neutrons produced by accelerator induced nuclear reactions. The neutron energy depends on the energy of the incident particle and nuclear reaction. Therefore, the BSA should be designed considering the neutron energy dependency. Now, several kinds of accelerators and reactions are used and proposed. Here, neutron producing reaction, moderator system, and activations are introduced.

Radiation safety analysis for the A-BNCT facility in Korea

A Proton Accelerator based Boron Neutron Capture Therapy (A-BNCT) facility is under development in Korea. Neutron beams for treatment are produced from a beryllium (Be) target and an 8 mA, 10 MeV proton beam. The purpose of the research is a radiation shielding analysis and an activation analysis for the facility design satisfying the radiation safety requirements as well as obtaining an operating license for the radiation facility according to a domestic nuclear commissioning procedure. The radiation shielding analysis was performed using the MCNPX computational particle transport code. The radiation source terms in the facility were evaluated and utilized in the shielding calculations. The minimum concrete thickness satisfying the designated dose rate of 5 μSv/h for the worker's area and 0.25 μSv/h for the public area were estimated and applied to the design. For an assessment of the radiation safety inside the facility, the dose rates were evaluated at several positions, such as behind the shielding door, around the primary barriers near the radiation sources, and in the penetrations of the ducts. The dose rate distribution was mapped for verification of the radiation safety for the entire facility. An activation analysis was carried out for the concrete walls, air, target assembly, beryllium target, and cooling water using FISPACT-2010 code. Concentrations of the activation products and dose rate induced by the radionuclides after shutdown were evaluated for the purpose of safe operation of the facility. The results were reviewed with the radiation safety regulations in Korea. As a result, it was proved that the final facility design satisfies the safety requirements.

Beam shaping assembly design of 7Li(p,n)7Be neutron source for boron neutron capture therapy of deep-seated tumor

The development of a medical facility for boron neutron capture therapy at Budker Institute of Nuclear Physics is under way. The neutron source is based on a tandem accelerator with vacuum insulation and lithium target. The proposed accelerator is conceived to deliver a proton beam around 10 mA at 2.3 MeV proton beam.To deliver a therapeutic beam for treatment of deep-seated tumors a typical Beam Shaping Assembly (BSA) based on the source specifications has been explored. In this article, an optimized BSA based on the 7Li(p,n)7Be neutron production reaction is proposed.To evaluate the performance of the designed beam in a phantom, the parameters and the dose profiles in tissues due to the irradiation have been considered.In the simulations, we considered a proton energy of 2.3 MeV, a current of 10 mA, and boron concentrations in tumor, healthy tissues and skin of 52.5 ppm, 15 ppm and 22.5 ppm, respectively. It is found that, for a maximum punctual healthy tissue dose seated to 11 RBE-Gy, a mean dose of 56.5 RBE Gy with a minimum of 52.2 RBE Gy can be delivered to a tumor in 40 min, where the therapeutic ratio is estimated to 5.38.All of these calculations were carried out using the Monte Carlo MCNP code.

SCOPING ANALYSIS OF MATERIAL ACTIVATION AT A BORON NEUTRON CAPTURE THERAPY FACILITY BASED ON THE Be(p,xn) REACTION WITH 30 MeV PROTONS.

Material activation assessment of a proposed accelerator-based boron neutron capture therapy facility was performed using the FLUKA Monte Carlo code to quantify the magnitude of the problem in terms of the isotope inventories, induced activities, and residual dose rates. Two simplified operation scenarios were considered: a 30-min proton bombardment to simulate a typical session of patient treatment and a long-term 1 year continuous operation to estimate the accumulation of long-lived radionuclides. Following the generation and transport of decay radiation, the space- and time-dependent inventories of induced radionuclides in materials and residual dose rates after shutdown were obtained. The predicted results were compared with the corresponding regulatory limits. Moreover, the effectiveness of various measures to reduce the impact of material activation was demonstrated.

Conceptual design of BNCT facility based on the TRR medical room

This paper presents a conceptual design of the Boron Neutron Capture Therapy (BNCT) facility based on the medical room of Tehran Research Reactor (TRR). The medical room is located behind the east wall of the reactor pool. The designed beam line is an in-pool Beam Shaping Assembly (BSA) which is considered between the reactor core and the medical room wall. The final designed BSA can provide 2.96× 109 n/cm2⋅s epithermal neutron flux at the irradiation position with acceptable beam contamination to use as a clinical BNCT.

Internal Dose Analysis for Radiation Worker in Cancer Therapy Based on Boron Neutron Capture Therapy with Neutron Source Cyclotron 30 MeV Using Monte Carlo N Particle Extended Simulator

Based Studies were carried out to analyze internal dose for radiation worker at Boron Neutron Capture Therapy (BNCT) facility base on Cyclotron 30 MeV with BSA and room that actually design before. This internal dose analyze include interaction between neutron and air. The air contains N2 (72%), O<sub>2 </sub>(20%), Ar (0.93%), CO<sub>2</sub>, Neon, Kripton, Xenon, Helium and Methane. That internal dose to the worker should be bellow limit dose for radiation worker amount of 20 mSv/years. From the particle that are present in the air, only Nitrogen and Argon can change into radioactive element. Nitrogen-14 activated to Carbon-14, Nitrogen-15 activated to Nitrogen-16, and Argon-40 activated to Argon-41. Calculation using tally facility in Monte Carlo N Particle Version Extended (MCNPX) program for calculated flux Neutron in the air 3,16x10<sup>7</sup> Neutron/cm<sup>2</sup>s. room design in cancer facility have a measurement of length 200 cm, width 200 cm and high 166,40 cm. flux neutron can be used to calculated the reaction rate which is 80,1x10<sup>-2</sup> reaction/cm<sup>3</sup>s for carbon-14 and 8,75x10<sup>-5</sup> reaction/cm<sup>3</sup>s. Internal dose exposed to the radiation worker is 9.08E-9 µSv.