Characterization and clinical utility of different collimator shapes in accelerator-based BNCT systems for head and neck cancer.

NeuCure® is the only accelerator-based boron neutron capture therapy (BNCT) system in the world with pharmaceutical approval. Until now, only flat collimators (FCs) on the patient side have been installed. However, in some cases of head and neck cancer patients, positioning the patient close enough to the collimator when using FCs was difficult. Thus, there are concerns about the prolongation of the irradiation time and overdose to normal tissues. To address these issues, a collimator with a convex-extended section on the patient side (extended collimators [ECs]) was developed, and its pharmaceutical approval was obtained in February 2022. This study evaluated the physical characterization and usefulness of each collimator using a simple geometry water phantom model and human model. In the water phantom model, the thermal neutron fluxes at 2 cm depth on the central axis were 5.13 × 108, 6.79 × 108, 1.02 × 109, and 1.17 × 109n/cm2/s for FC(120), FC(150), EC50(120), and EC100(120), respectively, when the distance from the irradiation aperture was kept constant at 18 cm. With ECs, the relative off-axis thermal neutron flux decreased steeply. In the hypopharyngeal cancer human model, the tumor dose changes were within <2%, but the maximum oral mucosa doses were 7.79, 8.51, 6.76, and 4.57 Gy-Eq, respectively. The irradiation times were 54.3, 41.3, 29.2, and 24.8 min, respectively. In cases where positioning the patient close to the collimator is difficult, the use of ECs may reduce the dose to normal tissues and shorten the irradiation time.

Thermal neutron detection and track recognition method in reference and out-of-field radiotherapy FLASH electron fields using Timepix3 detectors.

Objective.This work presents a method for enhanced detection, imaging, and measurement of the thermal neutron flux.Approach. Measurements were performed in a water tank, while the detector is positioned out-of-field of a 20 MeV ultra-high pulse dose rate electron beam. A semiconductor pixel detector Timepix3 with a silicon sensor partially covered by a6LiF neutron converter was used to measure the flux, spatial, and time characteristics of the neutron field. To provide absolute measurements of thermal neutron flux, the detection efficiency calibration of the detectors was performed in a reference thermal neutron field. Neutron signals are recognized and discriminated against other particles such as gamma rays and x-rays. This is achieved by the resolving power of the pixel detector using machine learning algorithms and high-resolution pattern recognition analysis of the high-energy tracks created by thermal neutron interactions in the converter.Main results. The resulting thermal neutrons equivalent dose was obtained using conversion factor (2.13(10) pSv·cm2) from thermal neutron fluence to thermal neutron equivalent dose obtained by Monte Carlo simulations. The calibrated detectors were used to characterize scattered radiation created by electron beams. The results at 12.0 cm depth in the beam axis inside of the water for a delivered dose per pulse of 1.85 Gy (pulse length of 2.4µs) at the reference depth, showed a contribution of flux of 4.07(8) × 103particles·cm-2·s-1and equivalent dose of 1.73(3) nSv per pulse, which is lower by ∼9 orders of magnitude than the delivered dose.Significance. The presented methodology for in-water measurements and identification of characteristic thermal neutrons tracks serves for the selective quantification of equivalent dose made by thermal neutrons in out-of-field particle therapy.

Sub-ps Pulsed Laser Deposition of Boron Films for Neutron Detector Applications.

In view of the demand for high-quality thermal neutron detectors, boron films have recently attracted widespread research interest because of their special properties. In this work, we report on the deposition of boron films on silicon substrates by sub-picosecond pulsed laser deposition (PLD) at room temperature. Particular emphasis was placed on the investigation of the effect of the laser energy density (fluence) on the ablation process of the target material, as well as on the morphological properties of the resulting films. In addition, based on the study of the ablation and deposition rates as a function of the fluence, the ablation/deposition mechanisms are discussed. We show that well-adherent and stable boron films, with good quality surfaces revealing a good surface flatness and absence of cracks, can be obtained by means of the PLD technique, which proves to be a reliable and reproducible method for the fabrication of thick boron coatings that are suitable for neutron detection technology.

Microdosimetry of an accelerator based thermal neutron field for Boron Neutron Capture Therapy.

The MUNES project (MUltidisciplinary NEutron Source) aims at the realization of an intense accelerator-based source of thermal neutrons, suitable for Boron Neutron Capture Therapy (BNCT). This exploits the interaction of 5 MeV protons onto a beryllium target, producing a fast neutron spectrum, which is moderated to the thermal range by a large assembly made of a Polytetrafluoroethylene (PTFE) tank filled with heavy water, surrounded by graphite blocks. The thermal neutron field is extracted through a bismuth beam port. The microdosimetric characterization of this field was performed using a cylindrical avalanche-confinement Tissue Equivalent Proportional Counter (TEPC) equipped with interchangeable cathode walls, positioned in front of the beam port. Measurements were taken both with a boron-doped wall and with an undoped one. The comparison of the two microdosimetric distributions allows to distinguish the relative dose contribution due to alpha particles and lithium ions from the BNC reaction from that of photons and other particles from neutron interactions on the cathode walls. The Relative Biological Effectiveness (RBE) was also calculated from the convolution of the measured spectra with a biological weighting function. This paper describes the experimental microdosimetric approach and the results of measurements with a boron-loaded cathode performed for the first time at an accelerator-based BNCT source.

Extending MIEZE spectroscopy towards thermal wavelengths.

A modulation of intensity with zero effort (MIEZE) setup is proposed for high-resolution neutron spectroscopy at momentum transfers up to 3 Å-1, energy transfers up to 20 meV and an energy resolution in the microelectronvolt range using both thermal and cold neutrons. MIEZE has two prominent advantages compared with classical neutron spin echo. The first is the possibility to investigate spin-depolarizing samples or samples in strong magnetic fields without loss of signal amplitude and intensity. This allows for the study of spin fluctuations in ferromagnets, and facilitates the study of samples with strong spin-incoherent scattering. The second advantage is that multi-analyzer setups can be implemented with comparatively little effort. The use of thermal neutrons increases the range of validity of the spin-echo approximation towards shorter spin-echo times. In turn, the thermal MIEZE option for greater ranges (TIGER) closes the gap between classical neutron spin-echo spectroscopy and conventional high-resolution neutron spectroscopy techniques such as triple-axis, time-of-flight and back-scattering. To illustrate the feasibility of TIGER, this paper presents the details of its implementation at the RESEDA beamline at FRM II by means of an additional velocity selector, polarizer and analyzer.

Design of a filtration system to improve the dose distribution of an accelerator-based neutron capture therapy system.

PURPOSE: The aim of this study is to design and evaluate a neutron filtration system to improve the dose distribution of an accelerator-based neutron capture therapy system. METHODS: An LiF-sintered plate composed of 99%-enriched 6 Li was utilized to filter out low-energy neutrons to increase the average neutron energy at the beam exit. A 5-mm thick filter to fit inside a 12-cm diameter circular collimator was manufactured, and experimental measurements were performed to measure the thermal neutron flux and gamma-ray dose rate inside a water phantom. The experimental measurements were compared with the Monte Carlo simulation, particle, and heavy ion transport code system. Following the experimental verification, three filter designs were modeled, and the thermal neutron flux and the biologically weighted dose distribution inside a phantom were simulated. Following the phantom simulation, a dummy patient CT dataset was used to simulate a boron neutron capture therapy (BNCT) irradiation of the brain. A mock tumor located at 4, 6, 8 cm along the central axis and 4-cm off-axis was set, and the dose distribution was simulated for a maximum total biologically weighted brain dose of 12.5 Gy with a beam entering from the vertex. RESULTS: All three filters improved the beam penetration of the accelerator-based neutron source. Filter design C was found to be the most suitable filter, increasing the advantage depth from 9.1 to 9.9 cm. Compared with the unfiltered beam, the mean weighted dose in the tumor located at a depth of 8 cm along the beam axis was increased by ∼25%, and 34% for the tumor located at a depth of 8 cm and off-axis by 4 cm. CONCLUSION: A neutron filtration system for an accelerator-based BNCT system was investigated using Monte Carlo simulation. The proposed filter design significantly improved the dose distribution for the treatment of deep targets in the brain.

Study of the 199Au nanoparticles production parameters via irradiation of platinum target by using thermal neutrons.

In this study, the production parameters of 199Au nanoparticles (199AuNPs) have been investigated by a two-part study. The first part is about investigating the indirect method of producing non-carrier-added (NCA) 199Au radionuclide. MCNPX-2.6, TALYS-1.9, and ALICE/ASH-0.1 codes were applied as the theoretical approach to simulate the core of Tehran research reactor (TRR) for determining the activity of 199Au, specifying the production yield of 199Au, and calculating the excitation function of P198t(n,γ)P199t→A199u reaction. As the corresponding experimental approach, two 11 mg and 15.5 mg samples of enriched 198Pt metal powder were irradiated by thermal neutrons for 21 h and 10 min. The liquid-liquid extraction (LLX) technique has been used with a different solvent for each sample. LLX using ethyl acetate and LLX using Di-(2-Ethylhexyl) phosphoric acid (HDEHP) were applied for the 15.5 mg and the 11 mg samples respectively. The chemical yield of 199Au was calculated more than %99 for the 15.5 mg sample, and more than %80 for the 11 mg sample. The second part is about synthesizing 199AuNPs in an average size of 50 nm by using Turkevich method.

Utilization of thermal neutron induced in-situ chain reactions and the (n,p) reaction with fast neutrons for compositional characterization of lithium titanate.

A methodology has been developed for the complete compositional characterization of lithium titanate (LTO) using neutron activation, which is quite challenging and no literature report is available so far. The concept of thermal neutron induced in-situ chain reactions 6Li(n,α)3H and 16O(3H,n)18F has been used for the determination of Li and O through the measurement of 18F activity. The method is capable of analyzing Li and O in percentage level as reported in the present analysis of two types of lithium titanate samples. Spectroscopic interference of the elements which can directly or indirectly affect the outcome, were evaluated meticulously. Determination of Ti was carried out using fast neutron activation through the product isotopes like 47Sc, 48Sc, generated via (n,p) nuclear reactions. Fast neutron activation methodology seems to be advantageous for Ti determination over thermal neutron activation, as it offers self validation through different isotopes and multiple gamma lines.

Development of a dose distribution shifter to fit inside the collimator of a Boron Neutron Capture Therapy irradiation system to treat superficial tumours.

The Kansai BNCT Medical Center has a cyclotron based epithermal neutron source for clinical Boron Neutron Capture Therapy. The system accelerates a proton to an energy of 30 MeV which strikes a beryllium target producing fast neutrons which are moderated down to epithermal neutrons for BNCT use. While clinical studies in the past have shown BNCT to be highly effective for malignant melanoma of the skin, to apply BNCT for superficial lesions using this system it is necessary to shift the thermal neutron distribution so that the maximum dose occurs near the surface. A dose distribution shifter was designed to fit inside the collimator to further moderate the neutrons to increase the surface dose and reduce the dose to the underlying normal tissue. Pure polyethylene was selected, and a Monte Carlo simulation was performed to determine the optimum thickness of the polyethylene slab. Compared with the original neutron beam, the shifter increased the thermal neutron flux at the skin by approximately 4 times. The measured and simulated central axis depth distribution and off axis distribution of the thermal neutron flux were found to be in good agreement. Compared with a 2 cm thick water equivalent bolus, a 26% increase in the thermal neutron flux at the surface was obtained, which would reduce the treatment time by approximately 29%. The DDS is a safe, simple and an effective tool for the treatment of superficial tumours for BNCT if an initially fast neutron beam requires moderation to maximise the thermal neutron flux at the tissue surface.

A thermal neutron source for the CERN radiation Calibration Laboratory.

This work presents the design, construction and experimental characterisation of a lightweight and low-cost thermal neutron assembly, to be used with the existing Am-Be source irradiator of CERN radiation Calibration Laboratory (Cal Lab). The assembly consists of a cylindrical moderator (18 cm diameter, 25.5 cm height and 5.5 kg weight) and an optional reflector box (5 cm thick walls, 20 kg weight). The moderator is tailored to fit on the Am-Be source in its irradiation position, while the box encloses the detector under test during the irradiation. The exposure volume delimited by the box is 30 × 30 × 30 cm3. The thermal neutron fluence at the exposure location, i.e., 30 cm from the source, was optimized by FLUKA Monte Carlo (MC) simulations. The simulations were validated with measurements performed with a bare 3He proportional counter. The thermal neutron fluence at the nominal irradiation position is 7.43 × 102 cm-2s-1 with the cylindrical moderator only, and 5.75 × 103 cm-2s-1 with the cylinder and the reflector box, with the detector placed at the centre of the box. The thermal neutron fluence inside the box is rather uniform (variation <5%).

TENIS - ThErmal Neutron Imaging System for use in BNCT.

A thermal neutron flux measurement tool with two perpendicular sets of plastic scintillator arrays was designed and simulated (Ghal-Eh and Green, 2016) with the MCNPX code (Version 2.6.0, with ENDF/B-VII cross section library (ENDF, 2011)). The proposed system aimed to provide a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions. In the present work, using Monte Carlo code FLUKA and its scintillation light transport capability, several important upgrades were carried out to include the light transport modeling in the response of plastic scintillators, analyze the cross-talk phenomenon, optimize the system geometry, and also provide a new approach in thermal neutron image reconstruction. The results showed that the last two cases played a significant role in improving the longitudinal profile of thermal neutron flux.

Thermal Neutron Relative Biological Effectiveness Factors for Boron Neutron Capture Therapy from In Vitro Irradiations.

The experimental determination of the relative biological effectiveness of thermal neutron factors is fundamental in Boron Neutron Capture Therapy. The present values have been obtained while using mixed beams that consist of both neutrons and photons of various energies. A common weighting factor has been used for both thermal and fast neutron doses, although such an approach has been questioned. At the nuclear reactor of the Institut Laue-Langevin a pure low-energy neutron beam has been used to determine thermal neutron relative biological effectiveness factors. Different cancer cell lines, which correspond to glioblastoma, melanoma, and head and neck squamous cell carcinoma, and non-tumor cell lines (lung fibroblast and embryonic kidney), have been irradiated while using an experimental arrangement designed to minimize neutron-induced secondary gamma radiation. Additionally, the cells were irradiated with photons at a medical linear accelerator, providing reference data for comparison with that from neutron irradiation. The survival and proliferation were studied after irradiation, yielding the Relative Biological Effectiveness that corresponds to the damage of thermal neutrons for the different tissue types.

Measurement of thermal neutrons reflection coefficients for two-layer reflectors.

In this research, thermal neutrons albedo coefficients and relative number of excess counts have been measured experimentally for different thicknesses of two-layer reflectors by using 241Am-Be neutron source (5.2Ci) and BF3 detector. Our used reflectors consist of two-layer which are combinations of water, graphite, polyethylene, and lead materials. Experimental results reveal that thermal neutron reflection coefficients slightly increased by addition of the second layer. The maximum value of growth for thermal neutrons albedo is obtained for lead-polyethylene compound (0.72 ± 0.01). Also, there is suitable agreement between the experimental values and simulation results by using MCNPX code.

Neutron reflectometry with the Multi-Blade 10B-based detector.

The Multi-Blade is a boron-10-based gaseous detector developed for neutron reflectometry instruments at the European Spallation Source in Sweden. The main challenges for neutron reflectometry detectors are the instantaneous counting rate and spatial resolution. The Multi-Blade has been tested on the CRISP reflectometer at the ISIS Neutron and Muon Source in the UK. A campaign of scientific measurements has been performed to study the Multi-Blade response in real instrumental conditions. The results of these tests are discussed in this paper.

Experimental and Monte Carlo investigation of mass attenuation coefficients of fission product isotopes in molecular precipitates.

Mass attenuation coefficients for molecular precipitates have been measured and compared to results using the EGS5 Monte Carlo computer code and results from an empirical formula. This study assesses the mass attenuation coefficients of isotopes that can be readily produced by thermal neutron activation of elements in a reactor and precipitated in molecular compounds. Good agreement exists between measured results and EGS5 simulated results. Results are within reasonable agreement with the empirical mass attenuation formula. These results are further compared to simulated fission product isotopes. This study suggests mass attenuation coefficients of molecular precipitates can be approximated using EGS5, especially in the instance of radioisotopes produced predominantly through uranium fission.

The thermal neutron facility HOTNES: theoretical design.

HOTNES (HOmogeneous Thermal NEutron Source) is a thermal neutron irradiation facility with extended and very uniform irradiation area. A 241Am-B radionuclide neutron source with nominal strenght 3.5×106 s-1 is located on bottom of a large cylindrical cavity (30cm diameter, 70cm in height) delimited by polyethylene walls. The upper part of this volume (30cm diameter, 40cm in height) is used to irradiate samples. A polyethylene cylinder, acting as shadowing object, prevents fast neutrons to directly reach the irradiation volume. Indeed neutrons can only reach the irradiation volume after multiple scattering with the cavity walls. The facility was designed trough extensive calculations with MCNPX. Irradiation planes are disks with 30cm diameter, centred on the cavity axis, and parallel to the cavity bottom. The value of thermal fluence in a given irradiation plane is as uniform as 1-2%. The value of thermal fluence rate simply depends on the height from the cavity bottom. Values of thermal fluence rate in the range 700-1000cm-2s-1 are available, depending on the irradiation plane chosen. The fraction of thermal neutrons is in the order of 90%, also depending on the irradiation plane. The angular distribution of thermal neutrons is roughly isotropic. Taking advantage of the HOTNES design, even large devices can be uniformly irradiated. This work presents HOTNES's design and describes the neutron field in the irradiation volume in terms of spatial, energy and direction distributions.

Prompt Gamma-ray Activation Analysis for Certification of Sulfur in Fuel Oil SRMs.

A combination of cold neutron prompt gamma-ray activation analysis (CNPGAA) and thermal neutron (TN) PGAA was used to determine sulfur in fuel oils to develop a method to provide values for certification. CNPGAA was used to measure S/H mass ratios, and TNPGAA to measure hydrogen mass fractions. Measurements were combined to determine sulfur mass fractions (with expanded uncertainties) of 2.159 % ± 0.072 % for SRM 1622e, 0.7066 % ± 0.0120 % for SRM 1619b, and 0.1266 % ± 0.0030 % for SRM 1617b, in agreement with certified values. The results validate the method as suitable for certification of sulfur at mass fractions ≥ 0.1 %.

Using a Tandem Pelletron accelerator to produce a thermal neutron beam for detector testing purposes.

Active thermal neutron detectors are used in a wide range of measuring devices in medicine, industry and research. For many applications, the long-term stability of these devices is crucial, so that very well controlled neutron fields are needed to perform calibrations and repeatability tests. A way to achieve such reference neutron fields, relying on a 3 MV Tandem Pelletron accelerator available at the CNA (Seville, Spain), is reported here. This paper shows thermal neutron field production and reproducibility characteristics over few days.

A plastic scintillator-based 2D thermal neutron mapping system for use in BNCT studies.

In this study, a scintillator-based measurement instrument is proposed which is capable of measuring a two-dimensional map of thermal neutrons within a phantom based on the detection of 2.22MeV gamma rays generated via nth+H→D+γ reaction. The proposed instrument locates around a small rectangular water phantom (14cm×15cm×20cm) used in Birmingham BNCT facility. The whole system has been simulated using MCNPX 2.6. The results confirm that the thermal flux peaks somewhere between 2cm and 4cm distance from the system entrance which is in agreement with previous studies.

ETHERNES: A new design of radionuclide source-based thermal neutron facility with large homogeneity area.

A new thermal neutron irradiation facility based on an (241)Am-Be source embedded in a polyethylene moderator has been designed, and is called ETHERNES (Extended THERmal NEutron Source). The facility shows a large irradiation cavity (45 cm × 45 cm square section, 63 cm in height), which is separated from the source by means of a polyethylene sphere acting as shadowing object. Taking advantage of multiple scattering of neutrons with the walls of this cavity, the moderation process is especially effective and allows obtaining useful thermal fluence rates from 550 to 800 cm(-2) s(-1) with a source having nominal emission rate 5.7×10(6) s(-1). Irradiation planes parallel to the cavity bottom have been identified. The fluence rate across a given plane is as uniform as 3% (or better) in a disk with 30 cm (or higher) diameter. In practice, the value of thermal fluence rate simply depends on the height from the cavity bottom. The thermal neutron spectral fraction ranges from 77% up to 89%, depending on the irradiation plane. The angular distribution of thermal neutrons is roughly isotropic, with a slight prevalence of directions from bottom to top of the cavity. The mentioned characteristics are expected to be attractive for the scientific community involved in neutron metrology, neutron dosimetry and neutron detector testing.