Neutron Capture Therapy Facility Research Articles

Neutron spectral fluence measurements using a Bonner sphere spectrometer in the development of the iBNCT accelerator-based neutron source

The neutron spectral fluence of an accelerator-based neutron source facility for boron neutron capture therapy (BNCT) based on a proton linac and a beryllium target was evaluated by the unfolding method using a Bonner sphere spectrometer (BSS). A 3He-proportional-counter-based BSS was used with weak beam during the development of the facility. The measured epithermal neutron spectra were consistent with calculations. The epithermal neutron intensity at the beam port was estimated and the results gave a numerical target for the enhancement of the proton beam intensity and will be used as reference data for measurements performed after the completion of the facility.

Optimization of beam shaping assembly based on D-T neutron generator and dose evaluation for BNCT

The feasibility of developing an epithermal neutron beam for a boron neutron capture therapy (BNCT) facility based on a high intensity D-T fusion neutron generator (HINEG) and using the Monte Carlo code SuperMC (Super Monte Carlo simulation program for nuclear and radiation process) is proposed in this study. The Monte Carlo code SuperMC is used to determine and optimize the final configuration of the beam shaping assembly (BSA). The optimal BSA design in a cylindrical geometry which consists of a natural uranium sphere (14 cm) as a neutron multiplier, AlF3 and TiF3 as moderators (20 cm each), Cd (1 mm) as a thermal neutron filter, Bi (5 cm) as a gamma shield, and Pb as a reflector and collimator to guide neutrons towards the exit window. The epithermal neutron beam flux of the proposed model is 5.73 × 109 n/cm2s, and other dosimetric parameters for the BNCT reported by IAEA-TECDOC-1223 have been verified. The phantom dose analysis shows that the designed BSA is accurate, efficient and suitable for BNCT applications. Thus, the Monte Carlo code SuperMC is concluded to be capable of simulating the BSA and the dose calculation for BNCT, and high epithermal flux can be achieved using proposed BSA.

Conceptual Design of Collimator at Boron Neutron Capture Therapy Facility with 30 MeV Cyclotron and Target 9Be as Neutron Generator Using Monte Carlo N-Particle Extended Simulator

The optimization of collimator has been studied which resulted epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) using Monte Carlo N Particle Extended (MCNPX). Cyclotron 30 MeV and <sup>9</sup>Be target is used as a neutron generator. The design criteria were based on recommendation from IAEA. Mcnpx calculations indicated by using 25 cm and 40 cm thickness of PbF<sub>2</sub> as reflector and back reflector, 15 cm thickness of TiF<sub>3</sub> as first moderator, 35 cm thickness of AlF<sub>3</sub> as second moderator, 25 cm thickness of <sup>60</sup>Ni as neutron filter, 2 cm thickness of Bi as gamma filter, and aperture with 20 cm of diameter size, an epithermal neutron beam with an intensity 1.21 × 10<sup>9</sup> n.cm<sup>-2</sup>.s<sup>-1</sup>, fast neutron and gamma doses per epithermal neutron of 7.04 × 10<sup>-13</sup> Gy.cm<sup>2</sup>.n<sup>-1</sup> and 1.61 × 10<sup>-13</sup> Gy.cm<sup>2</sup>.n<sup>-1</sup>, minimum thermal neutron per epithermal neutron ratio of 0.043, and maximum directionality of 0.58, respectively could be produced. The results have not passed all the IAEA’s criteria in fast neutron component and directionality.

Shielding analyses of an AB-BNCT facility using Monte Carlo simulations and simplified methods

Accurate Monte Carlo simulations and simplified methods were used to investigate the shielding requirements of a hypothetical accelerator-based boron neutron capture therapy (AB-BNCT) facility that included an accelerator room and a patient treatment room. The epithermal neutron beam for BNCT purpose was generated by coupling a neutron production target with a specially designed beam shaping assembly (BSA), which was embedded in the partition wall between the two rooms. Neutrons were produced from a beryllium target bombarded by 1-mA 30-MeV protons. The MCNP6-generated surface sources around all the exterior surfaces of the BSA were established to facilitate repeated Monte Carlo shielding calculations. In addition, three simplified models based on a point-source line-of-sight approximation were developed and their predictions were compared with the reference Monte Carlo results. The comparison determined which model resulted in better dose estimation, forming the basis of future design activities for the first ABBNCT facility in Taiwan.

An accelerator-based Boron Neutron Capture Therapy (BNCT) facility based on the 7Li(p,n)7Be

BNCT (Boron Neutron Capture Therapy) is a therapeutic modality used to irradiate tumors cells previously loaded with the stable isotope 10B, with thermal or epithermal neutrons. This technique is capable of delivering a high dose to the tumor cells while the healthy surrounding tissue receive a much lower dose depending on the 10B biodistribution. In this study, therapeutic gain and tumor dose per target power, as parameters to evaluate the treatment quality, were calculated. The common neutron-producing reaction 7Li(p,n)7Be for accelerator-based BNCT, having a reaction threshold of 1880.4keV, was considered as the primary source of neutrons. Energies near the reaction threshold for deep-seated brain tumors were employed. These calculations were performed with the Monte Carlo N-Particle (MCNP) code. A simple but effective beam shaping assembly (BSA) was calculated producing a high therapeutic gain compared to previously proposed facilities with the same nuclear reaction.

Dosimetry of in vitro and in vivo Trials in Thermal Column Kartini Reactor for Boron Neutron Capture Therapy (BNCT) facility by using MCNPX Simulator Code

<span>A dosimetry study of in vitro and in vivo trials system in thermal column of Kartini Reactor for Boron Neutron Capture Therapy (BNCT) facility has been conducted by using the Monte Carlo N-Particle Extended (MCNPX) software. Source of neutron originated from the 100 kW reactor which has been modified by the previous researcher. Models have been made by using simple geometries to represent tissues. Models of in vitro have been made by 4 spheres each has 1 cm diameter to represent tumour, whereas in vivo by 4 cylinders each has 6 cm length, 3 cm diameter, and breast soft tissue material with 1 cm sphere each located in the center of the cylinders to represent models of mouse with tumour. An increase in value of the boron concentration will increase the value of dose rate as well, then the exposure time should be shorter. The exposure times (in minutes) of in vitro trials for 20, 25, 30, 50, 75, 100, 125, and 150 μg boron/g tissues are 117.77, 117.77, 117.07, 115.69, 114.02, 112.39, 110.80, and 109.27. Whereas the exposure times of in vivo trials are 163.58, 162.78, 161.98, 158.88, 155.16, 151.61, 148.22, dan 144.98. In vitro trials have greater values of dose rate so that in vitro trials have shorter exposure time.</span>

Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

Studies were carried out to design a shielding for BNCT facility in the end of Kartini reactor’s thermal column with predesigned collimator. The design consist of selecting the material and their thickness. The shielding is required to absorb the leaking radiation until the Dose Limit Value of 20 mSv/year for radiation worker is met. The material considered were paraffin, barite concrete, borated polyethylene, stainless steel 304 and lead. The calculation was done using MCNPX tally facility with converted dose limit value of 10.42 µSv/hour. Design number two were chosen as the best from three designs which surrounded a room with length, width and height of, respectively 200 cm, 200 cm and 166.4 cm. The first and main layer are borated polyethyelene and barite concrete of 20 and 30 cm, respectively. The additional layer are borated polyethyelene and barite concrete of 15 cm and 15 cm with less volume than the main layer to decrease the primary straight radiation from the thermal column. Maximum radiation dose rate is 7.0746 µ Sv/hour in cell 227 with average dose rate of 2.58712 µSv/hour.

Characteristics and application of spherical-type activation detectors in neutron spectrum measurements at a boron neutron capture therapy (BNCT) facility

A set of spherical-type activation detectors was developed aiming to provide better determination of the neutron spectrum at the Tsing Hua Open-pool Reactor (THOR) BNCT facility. An activation foil embedded in a specially designed spherical holder exhibits three advantages: (1) minimizing the effect of neutron angular dependence, (2) creating response functions with broadened coverage of neutron energies by introducing additional moderators or absorbers to the central activation foil, and (3) reducing irradiation time because of improved detection efficiencies to epithermal neutron beam. This paper presents the design concept and the calculated response functions of new detectors. Theoretical and experimental demonstrations of the performance of the detectors are provided through comparisons of the unfolded neutron spectra determined using this method and conventional multiple-foil activation techniques.

Optimization study for BNCT facility based on a DT neutron generator

Background : A Boron Neutron Capture Therapy (BNCT) facility, based on a DT neutron generator, with the final goal to find out a potenal, alternave, soluon to exisng BNCT treatment facilies which a re based on nuclear reactors is examined. Materials and Method s: With the aim of the MCNP4B Monte Carlo code different beam-shaping assembly (BSA) configuraons were considered. Lead was selected as reflector material while CF 2, D 2O, Fluental, PbF 4, PbF 2, BiF 3, BiF 5, MgF 2, Al 2O3, AlF 3, TiF 3, BeD 2, CaF 2 and 7 LiF were examined as spectrum shi3ers. In order to improve t he quality of the beam �tanium, nickel-60, iron andtanium alloy (Ti 6Al 14 V) were simulated as fast neutrons filters while lead and bismuth were considered as gamma filters. Results: An extensive set of calculaons performed with MCNP4B Monte Carlo code have shown that the combinaon of 7 LiF which accommodates a conic part made of D 2O, then followed by a TiF 3 layer is the opmum moderator design. The use of three different materials for further reducon of fast neutrons, thermal neutrons and gamma rays is necessary. 60

A prospective study to assess the performance of the improved Boron Neutron Capture Therapy Facility in Argentina

From 2008 to 2011, several planned modifications were implemented at the RA-6 reactor in Argentina, leading to significant benefits for future BNCT treatments. New capabilities have been implemented in NCTPlan treatment planning system. To assess the performance of the new BNCT facility, a dosimetric reevaluation of previous clinical cases was performed, taking into account the modifications carried out in the new facility and compared the results of the original treatment plans with optimized plans that are considered as feasible patient setups.

Treatment planning capability assessment of a beam shaping assembly for accelerator-based BNCT

Within the frame of an ongoing project to develop a folded Tandem-Electrostatic-Quadrupole accelerator facility for Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT) a theoretical study was performed to assess the treatment planning capability of different configurations of an optimized beam shaping assembly for such a facility. In particular this study aims at evaluating treatment plans for a clinical case of Glioblastoma.

Feasibility study on the use of uranium in photoneutron target and BSA optimization for Linac based BNCT

A hybrid photoneutron target including natural uranium has been studied for a 20 MeV linear electron accelerator (Linac) based Boron Neutron Capture Therapy (BNCT) facility. In this study the possibility of using uranium to increase the neutron intensity has been investigated by focusing on the time dependence behavior of the build-up and decay of the delayed gamma rays from fission fragments and activation products through photo-fission reactions in the BSA (Beam Shaping Assembly) configuration design. Delayed components of neutrons and photons were calculated. The obtained BSA parameters are in agreement with the IAEA recommendation and compared to the hybrid photoneutron target without U. The epithermal flux in the suggested design is 2.67E9 (n/cm 2s/mA).

BINP accelerator based epithermal neutron source

Innovative facility for neutron capture therapy has been built at BINP. This facility is based on compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915-2.5 MeV protons bombarding a lithium target using (7)Li(p,n)(7)Be threshold reaction. In the article, diagnostic techniques for proton beam and neutrons developed are described, results of experiments on proton beam transport and neutron generation are shown, discussed, and plans are presented.

Development of high intensity ion sources for a Tandem-Electrostatic-Quadrupole facility for Accelerator-Based Boron Neutron Capture Therapy

Several ion sources have been developed and an ion source test stand has been mounted for the first stage of a Tandem-Electrostatic-Quadrupole facility For Accelerator-Based Boron Neutron Capture Therapy. A first source, designed, fabricated and tested is a dual chamber, filament driven and magnetically compressed volume plasma proton ion source. A 4 mA beam has been accelerated and transported into the suppressed Faraday cup. Extensive simulations of the sources have been performed using both 2D and 3D self-consistent codes.

Development of a Tandem-Electrostatic-Quadrupole facility for Accelerator-Based Boron Neutron Capture Therapy

We describe the present status of an ongoing project to develop a Tandem-ElectroStatic-Quadrupole (TESQ) accelerator facility for Accelerator-Based (AB)-BNCT. The project final goal is a machine capable of delivering 30 mA of 2.4 MeV protons to be used in conjunction with a neutron production target based on the (7)Li(p,n)(7)Be reaction. The machine currently being constructed is a folded TESQ with a high-voltage terminal at 0.6 MV. We report here on the progress achieved in a number of different areas.

Electrostatic design and beam transport for a folded tandem electrostatic quadrupole accelerator facility for accelerator-based boron neutron capture therapy

Within the frame of an ongoing project to develop a folded Tandem-Electrostatic-Quadrupole (TESQ) accelerator facility for Accelerator-Based Boron Neutron Capture Therapy (AB-BNCT), we discuss here the electrostatic design of the machine, including the accelerator tubes with electrostatic quadrupoles and the simulations for the transport and acceleration of a high intensity beam.

Building of scientific information system for sustainable development of BNCT in Bulgaria

Building a boron neutron capture therapy (BNCT) facility is foreseen within the reconstruction of the Research Reactor IRT (IRT) of the Institute for Nuclear Research and Nuclear Energy of the Bulgaria Academy of Sciences (INRNE). The development of BNCT at IRT plays a very significant role in the plan for sustainable application of the reactor. A centralized scientific information system on BNCT is being built at the INRNE with the purpose to collect and sort new information as knowledge accumulated during more than thirty years history of BNCT. This BNCT information system will help the creation and consolidation of a well informed and interconnected interdisciplinary team of physicists, chemists, biologists, and radio-oncologists for establishing BNCT cancer treatment in Bulgaria. It will strengthen more intensive development of the national network as well as its enlargement to the Balkan region countries. Furthermore, to acquaint the public at large with the opportunity for BNCT cancer treatment will be addressed. Human, social, and economics results due to BNCT for many patients from Balkan region are expected.

First neutron generation in the BINP accelerator based neutron source

Pilot innovative facility for neutron capture therapy was built at Budker Institute of Nuclear Physics, Novosibirsk. This facility is based on a compact vacuum insulation tandem accelerator designed to produce proton current up to 10 mA. Epithermal neutrons are proposed to be generated by 1.915 MeV protons bombarding a lithium target using 7Li(p,n) 7Be threshold reaction. The results of the first experiments on neutron generation are reported and discussed.

Angle and energy differential neutron spectrometry for the SPES BNCT facility

An accelerator-driven thermal neutron facility for boron neutron capture therapy of skin melanoma is currently under construction at the Legnaro National Laboratories, Italy. The installation relies on the production of neutrons from a thick beryllium target bombarded with 5 MeV protons. A complete set of double differential data, i.e. angle- and energy-differential neutron spectra produced by the beryllium target, is necessary for the Monte Carlo-based design of the installation. For this purpose, double differential fluence measurements are currently performed with the “BINS” neutron spectrometer using 5 MeV protons at the “CN” Van de Graaf accelerator. This spectrometer uses a superheated emulsion of dichlorotetrafluoroethane which is sequentially operated at 25, 30, 35, 40, 45, 50 and 55 °C and thus provides a series of seven sharp thresholds covering the 0.1–10 MeV neutron energy interval. Deconvolution of the data is performed with the code “MAXED”, which is based on the maximum entropy principle. The analysis of our first neutron spectrometry measurements at angles of 0°, 40°, 80° and 120° supports the viability of the BINS spectrometry method for the generation of the required double differential data.

Extension of the calibration curve for the PGRA facility in Petten

At the boron neutron capture therapy (BNCT) facility in Petten, the Netherlands, 10B concentrations in biological materials are measured with the prompt gamma ray analyses facility that is calibrated using certified 10B solutions ranging from 0 to 210 ppm. For this study, newly certified 10B solutions ranging up to 1972 ppm are added. MCNP simulations of the setup range to 5000 ppm. A second order polynomial (as already used) will fit 10B-concentrations less than 300 ppm. Above 300 ppm a fitted third order polynomial is needed to describe the calibration curve accurately.