Thermal Neutron Fluence Research Articles

Optimization study on the compact thermal neutron imaging system based on portable D-T neutron generator

The distribution of lithium in lithium-ion batteries is critical to their performance and lifespan. Studies have shown that thermal neutron imaging can effectively perform non-destructive testing of this distribution. However, most current thermal neutron imaging systems rely on expensive, bulky, and immobile nuclear reactors or spallation neutron sources, significantly limiting their application scenarios. Therefore, this study adopts a portable D-T neutron generator as the neutron source and uses Monte Carlo simulation methods to optimize the design of key components in the thermal neutron imaging system. Simulation results indicate that the optimized thermal neutron imaging system achieves a thermal neutron flux of 8.36 × 104 n·cm−2·s−1 and a thermal neutron content (TNC) of 14.4% at a collimation ratio of 50. Within an imaging range of Φ100 mm, the non-uniformity of the thermal neutron fluence is approximately 2.07%. Simulation imaging results of lithium and iron step wedges demonstrate that the compact thermal neutron imaging system designed in this study can effectively distinguish between iron and lithium of the same thickness. This finding provides an important reference for subsequent detection of lithium distribution within lithium-ion batteries.

Characteristics of a neutron source based on an electron accelerator LINAC-200

As part of the commissioning work at the linear electron accelerator LINAC-200 (JINR), we conducted experiments to study the characteristics of a pulsed neutron source obtained by irradiating a converter target with a 140 MeV electron beam. To estimate neutron source parameters and neutron energy spectra in lead and tungsten targets with different sizes numerical simulations were performed. Additionally, we have conducted experiments to measure the energy spectra of neutrons and gamma rays using high-resolution time-of-flight method. It has been found that the ratio of integral estimates between the calculated and average measured neutron yields (for tungsten and lead targets) did not exceed 30%. The fluence of resonant and thermal neutrons in the target is estimated at 2.8×1013 neutrons per second, which corresponds to the requirements for modern pulsed neutron sources.

Isotopic variations of Sm, Gd, Er and Yb found in planetary materials caused by neutron-capture reactions in nature

The isotopic shifts of 149Sm–150Sm and 157Gd–158Gd have often been observed in meteorites and lunar surface materials, because they result from the neutron-capture reactions associated with secondary neutrons produced by cosmic-ray irradiation. While the Sm and Gd isotopic shifts can mainly be used for the estimation of thermal neutron fluences that of 167Er–168Er has recently been applied in the estimation of epithermal neutron fluences. The systematic isotopic dataset of Sm, Gd and Er helps us to consider the details of planetary materials’ cosmic-ray exposure conditions using the balance of the fluences between thermal and epithermal neutrons. This paper reviews a series of isotopic variations of Sm, Gd, and Er in association with neutron-capture reactions for the application of planetary sciences. As a new attempt and possibility for better understanding the neutron fluence and its energy distribution, the use of Yb isotopic variation is then discussed using two different data sources, namely lunar regolith and the Oklo natural reactors. Finally, the preliminary result for the precise isotopic measurement of Yb is presented from the viewpoint of chemical separation and instrumental improvement.

Neutron Capture Enhances Dose and Reduces Cancer Cell Viability in and out of Beam During Helium and Carbon Ion Therapy

Neutron capture enhanced particle therapy (NCEPT) is a proposed augmentation of charged particle therapy that exploits thermal neutrons generated internally, within the treatment volume via nuclear fragmentation, to deliver a biochemically targeted radiation dose to cancer cells. This work is the first experimental demonstration of NCEPT, performed using both carbon and helium ion beams with 2 different targeted neutron capture agents (NCAs). Human glioblastoma cells (T98G) were irradiated by carbon and helium ion beams in the presence of NCAs [10B]-BPA and [157Gd]-DOTA-TPP. Cells were positioned within a polymethyl methacrylate phantom either laterally adjacent to or within a 100 × 100 × 60 mm spread out Bragg peak (SOBP). The effect of NCAs and location relative to the SOBP on the cells was measured by cell growth and survival assays in 6 independent experiments. Neutron fluence within the phantom was characterized by quantifying the neutron activation of gold foil. Cells placed inside the treatment volume reached 10% survival by 2 Gy of carbon or 2 to 3 Gy of helium in the presence of NCAs compared with 5 Gy of carbon and 7 Gy of helium with no NCA. Cells placed adjacent to the treatment volume showed a dose-dependent decrease in cell growth when treated with NCAs, reaching 10% survival by 6 Gy of carbon or helium (to the treatment volume), compared with no detectable effect on cells without NCA. The mean thermal neutron fluence at the center of the SOBP was approximately 2.2 × 109 n/cm2/Gy (relative biological effectiveness) for the carbon beam and 5.8 × 109 n/cm2/Gy (relative biological effectiveness) for the helium beam and gradually decreased in all directions. The addition of NCAs to cancer cells during carbon and helium beam irradiation has a measurable effect on cell survival and growth in vitro. Through the capture of internally generated neutrons, NCEPT introduces the concept of a biochemically targeted radiation dose to charged particle therapy. NCEPT enables the established pharmaceuticals and concepts of neutron capture therapy to be applied to a wider range of deeply situated and diffuse tumors, by targeting this dose to microinfiltrates and cells outside of defined treatment regions. These results also demonstrate the potential for NCEPT to provide an increased dose to tumor tissue within the treatment volume, with a reduction in radiation doses to off-target tissue.

A compact moderator for thermal neutron reference radiation field with an 241Am-Be source

To calibrate thermal neutron detectors, State Key Laboratory of NBC Protection for Civilian plans to establish thermal neutron reference radiation field (TNRRF) based on the existing 7.4 × 1011 Bq 241Am-Be source. An 82 cm (L) × 50 cm (W) × 50 cm (H) moderator was designed through Monte Carlo simulation to thermalize fast neutrons. Each component was determined to optimize thermal neutron fluence rate φ, proportion of thermal neutrons ρ, and uniformity factor μ of the radiation field. A 3He proportional counter was used to evaluate the fabricated moderator. Results show that surrounding reflector is the main origin of thermal neutrons. D2O tank can promote ρ and its length can be utilized to adjust φ. Homogenizing lens will improve μ and reduce φ. Experimental φ, cadmium ratio η, and μ at the reference position are 1808.82 ± 85.96 cm−2 s−1, 315.80, and 5.29 %, respectively, which are consistent with simulated results. Next, TNRRF will be established after long-term stability testing.

Studies of irradiated two-phase lithium ceramics Li4SiO4/Li2TiO3 by thermal desorption spectroscopy

Two-phase ceramics Li2TiO3-Li4SiO4 are one of the potentially promising materials for creating a ceramic blanket for DEMO reactor. However, until now, a limited number of studies have been carried out on the release of tritium from this material under the influence of neutron irradiation, which poses fundamental problems for its application.In this work, the pebbles of two-phase lithium ceramics consisting of 25 mol% Li2TiO3 and 75 mol% Li4SiO4 were studied using thermal desorption spectroscopy (TDS) after their irradiation with neutrons at the WWR-K research reactor. Under the conditions of the irradiation experiment, which lasted 21.5 days at a reactor power of 6 MW, the samples were exposed to thermal neutrons with a flux density of 2·1013n/(cm2 × s). The thermal neutron fluence accumulated as a result of irradiation was 3.7·1019 cm−2.It has been established that tritium comes out mainly in the form of HT molecules and has three visible TDS peaks, which can be described by three rates of the first-order desorption reaction with an activation energy of 90 kJ/mol. Experiments have shown that the main amount of tritium is evenly distributed throughout the pebble’s pores and voids. Its yield is determined by desorption and transfer from inner voids to the boundaries that communicate with the outer surface of the sample. The kinetics of tritium release depends on the number, size and depth of the exit paths of such areas inside the ceramic.

Activation level of the concrete building and pressure vessel in JAEA-Tokai tandem accelerator

ABSTRACT The activation level of the JAEA-Tokai tandem accelerator facility was investigated experimentally in advance of the future decommissioning. JAEA-Tokai tandem accelerator facility has a higher terminal voltage of 18 MV and a larger total floor area, Compared to other electrostatic accelerators with terminal voltage of 1 MV to 6 MV. Therefore, determination for ‘where,’ ‘what,’ and ‘how many’ nuclides are produced in the facility is crucial. Thermal neutrons generated by beam losses associated with accelerator operations contribute significantly to the activation of equipment and facilities. The accumulated activities of60Co and152Eu; the most considerable radionuclides at the decommissioning, can be deduced by the thermal neutron fluence rate during the accelerator operation. In this study, thermal neutron fluence measurement on the surface of the pressure vessel and concrete building was conducted with conventional methods using dosimeters and metal foil detectors as well as a new method using a portable γ-ray detector. The thermal neutron fluence in the facility during the accelerator operation ranges from 101 to 104 n/cm2/s. The sum of deduced activities of60Co and152Eu in 50 years is much lower than the clearance level of 0.1 Bq/g in all areas except in the irradiation room.

Multicentre investigation of neutron contamination at cardiac implantable electronic device (CIED) location due to high-energy photon beams using passive detectors and Monte Carlo simulations

Radiotherapy treatments involving LINACs operating at accelerating potentials >10 MV generate (photo)neutrons which deliver dose to patients also outside the target volume. This effect is particularly relevant for patients with cardiac implantable electronic devices (CIEDs), which can be damaged by the therapeutic irradiation. In the last few years, there has been a rising interest in this issue, and it seems that damage to CIEDs is primarily associated with the thermal component of the photoneutron field. In particular, a recent study led by Politecnico di Milano considered CIEDs from various manufacturers and showed that some of these devices can be damaged after an irradiation with a thermal neutron fluence of about 10^{9}hbox { cm}^{-2}. The present work results from a collaboration among Politecnico di Milano, the University of Pisa, the University of Trieste and three Italian hospitals located in Lucca, Trieste and Varese, respectively, and it is primarily aimed at evaluating the thermal neutron fluence in CIED region for some high-energy treatments delivered at 15 and 18 MV and to determine whether it is comparable to the critical value given above, which has been experimentally determined to be potentially harmful for CIEDs. Thermal neutron fluence was measured through CR-39 detectors and TLDs, which were housed inside a BOMAB-like phantom mimicking the patient’s trunk. The experimental sessions involved two models of LINAC, Varian Clinac DHX (Varese hospital) and Elekta Synergy (Lucca and Trieste hospitals). The experimental results show that the treatments considered in this study can lead to a thermal neutron fluence in the cardiac region comparable to the critical value. Furthermore, detailed Monte Carlo geometries for the facilities involved in this project were developed with the MCNP code (v. 6.2), and they were tested by comparing simulation results to measurements considering some benchmark irradiation plans. Bubble detectors were also employed for fast neutron fluence measurements to be compared to simulation outputs. These computational models stand out as promising tools for the investigations required in this work, and they can be used for further studies also extending their use to analogous facilities hosting the same models of LINACs.

A model for calculating the irradiation swelling of AgInCd absorber in nuclear control rods

The actual swelling of AgInCd absorber might exceed the predicted swelling value after years of service in pressurized water reactors, and the chemical and microstructural changes of AgInCd absorber induced by transmutation reactions are the main reason for the swelling acceleration of AgInCd absorber. In the present study, a model for calculating the irradiation swelling of AgInCd absorber in nuclear control rods is developed according to chemical and microstructural changes of AgInCd absorber. In this model, the chemical compositions of AgInCd absorber as a function of the thermal neutron fluence are firstly calculated, and then the volume of AgInCd absorber after irradiation is obtained on the basis of the crystallographic parameters of phases in the AgInCd absorber, and the irradiation swelling of AgInCd absorber is finally calculated. The crystallographic parameters can be obtained by preparing the simulated AgInCd alloys and fitting the experimental data. The model calculating results of irradiation swelling are in good agreement with the actual swelling data in literature. More importantly, the present model can well explain the EPRI results of the acceleration in the diametral swelling rate above 6–8 × 1020 n/cm2 and the decrease in the diametral swelling rate above about 2 × 1021 n/cm2.

Survivability of microstructure semiconductor neutron detectors (MSND) in a thermal neutron radiation environment

The performance of five representative microstructured semiconductor neutron detectors (MSND) were investigated after exposure to varying magnitudes of thermal neutrons fluences ranging from 107 cm−2 to 1012 n cm−2. The devices have 95% enriched microparticulate 6LiF granules backfilled in microscopic trenches within a Si pvn diode. Neutron interactions in the 6Li causes the emission of reaction products which ionize the trenched Si diode depletion regions. The average intrinsic thermal-neutron detection efficiency (ϵtn) of the five devices before irradiation measured as 24.09% ± 0.35% with an operating leakage current below 11 nA at a −3 V reverse bias condition. The observed threshold of performance deterioration was approximately 1010 n cm−2, resulting in an spectral shift in the reaction product spectrum to lower energy channels and a general increase in leakage current. Although damage was observed beyond a thermal-neutron fluence of 1010 n cm−2, the devices remained operable. Beyond a thermal-neutron fluence of 1012 n cm−2, the devices experienced severe damage and demonstrated poor detector performance. At the highest thermal neutron fluences tested, near 1012 n cm−2, the average ϵtn of the five devices decreased to 4.42% ± 0.10% with an average operating leakage current of 300 nA at a −3 V reverse bias condition. Performance degradation was attributed to reaction-product damage in the MSND pvn diode structure.

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.

Al5BO9:Tb3+: Synthesis, structural characterization and thermoluminescence dosimetry studies for high intensity thermal neutron beams

In this manuscript, gel-combustion synthesis of Al5BO9:xTb3+ (x = 0.0%, 0.05%, 0.1%, 0.2% and 1.5%) phosphor along with the detailed structural characterizations using a host of techniques such as XRD, SEM, EDS, IR, EXAFS and PL etc. has been reported. XRD analysis confirmed the phase pure formation and also the nanocrystalline nature of the materials. Detailed EXAFS analysis explained the oxidation state and the coordination behaviour of Tb3+ ion in the Al5BO9 matrix. PL study showed that the material can be a potential green light emitting phosphor. Thermoluminescence (TL) study revealed the presence of a wide shoulder peak and a broad dosimetry peak situating at 551 K. Net TL response from 10B and 11B enriched Al5BO9:0.1%Tb3+ thin pellets showed prolonged linearity within the thermal neutron fluence range of 3.2 × 1010 to 1.6 × 1011 n/cm2. Also, the maximum fading of the TL signal is only 9% during the storage period of 112 days. These indicates that the developed material can be a potential candidate for dosimetry applications involving high intensity slow neutron beams. The TL glow curve was deconvoluted which revealed the presence of five individual peaks whose kinetic parameters viz. activation energy (E), order of kinetics (b) and frequency factor (s) were evaluated using both glow curve deconvolution and Chen's peak shape method and the results were found to be in good agreement.

A Monte Carlo study of high-LET particle production in proton boron therapy

We simulate the production of high-LET particles from different boron nuclear reactions in proton therapy. This study is motivated by neutron capture-enhanced particle therapy (NCEPT) (Safavi-Naeini et al. in Sci Rep 8(1):1–13, 2018) and the recently proposed proton boron capture therapy (PBCT) (Cirrone et al. in Sci Rep 8(1):1–15, 2018), where both therapy types suggest to utilize the production of high-LET particle for enhancing the RBE of the treatment. From Monte Carlo simulations performed in TOPAS (Perl et al. in Med Phys 39(11):6818–6837, 2012), we find that NCEPT reaction is the more prominent reaction of the two with a factor of 4000 pm 700 more particles produced. The reason for this is the relative high fluence of thermal and epithermal neutrons (neutrons < 1 hbox { keV}) present in clinical relevant proton fields. The creation of these neutrons is shown to be highly influenced by the target phantom and proton field size. By varying the field and phantom size, it was found that the fluence of low-energy neutrons can increase/decrease two orders of magnitude.

Thermoluminescent properties of Tb3+-doped B2O3–Na2O–CaO–P2O5 glasses for neutron detection

We fabricated Tb3+-doped B2O3–Na2O–CaO–P2O5 glasses using 10B- or 11B-enriched raw materials to develop new thermoluminescent (TL) glass materials for neutron detection. We investigated TL properties of the glass materials after X-ray and thermal neutron irradiation. A TL glow peak was observed at 376 K after X-ray irradiation, and the integrated TL intensities of the 10B- and 11B-enriched glasses were similar. This result indicates that the 10B- and 11B-enriched glasses had similar sensitivities to X-rays. A broad peak was observed in the 350–650 K range for the 10B- and 11B-enriched glasses after neutron irradiation, and the 10B-enriched glass exhibited higher TL intensities than the 11B-enriched glass in the neutron fluence range of 1.0 × 105–1.0 × 1011 neutron cm−2. This result indicates that γ-rays and high-energy particles in the 10B(n, α)7Li reaction efficiently induced thermoluminescence in the 10B-enriched glass. The TL intensity derived from the 10B(n, α)7Li reaction linearly increased with the thermal neutron fluence in the neutron fluence range of 1.0 × 105–1.0 × 1011 neutron cm−2. The 10B- and 11B-enriched glasses exhibited a wider dynamic range and lower minimum detectable neutron fluence than quite a few TL materials for neutron detection. Therefore, TL glass materials for neutron detection were successfully developed.

Thermal neutron reference radiation facility with high thermalization and large uniformity area

A new thermal neutron facility based on 12 241Am–Be neutron sources has been established at the National Institute of Metrology, China. It has two independent irradiation fields, the inner field and outer field, constructed with high-purity graphite and heavy water as moderators, respectively. Three innovative designs, including the reflection cavity, reflection layer, and homogenizing lens, were introduced to improve the performance of the facility. The reflection layer can increase the thermal neutron fluence rate by about 3–4 times, the reflection cavity also enhances thermal fluence by 3–4 times by taking advantage of multiple scattering of neutrons, and the homogenizing lens mounted on the neutron exit surface of the moderator improve the distribution of thermal neutron. The characteristics of the two fields were investigated by Monte Carlo simulations and experimental measurements. The results show that the new facility has high thermalization and large uniformity area with these innovative designs. For the inner field, the thermal neutron fluence rate at the reference position is (21 433.3 ± 407.2) cm−2 s−1, the cadmium ratio is 20.6, the non-uniformity is 0.1% over the 40 cm × 40 cm region. For the outer field, the thermal fluence rate at the reference position is (2046.0 ± 49.1) cm−2 s−1, the cadmium ratio is 1433, and the non-uniformity is 0.7% in the 70 cm × 70 cm region. The facility is expected to be suitable for the large-size detector, and its excellent performance is attractive for various thermal neutron experiments.

Measuring the thermal neutron fluence of NTD-Ge using the self-monitoring method

Thermal neutron irradiation in reactors is widely used in the different frontier fields, including advanced materials and semiconductor devices. The total thermal neutron fluence is a key parameter in various applications. In this paper, we demonstrate a method to determine the total thermal neutron fluence during the manufacture of Neutron Transmutation Doping Germanium (NTD-Ge). The NTD-Ge samples are irradiated at the China Advanced Research Reactor, and their total neutron fluences are measured using the activity of radionuclide 71Ge through the reaction process 70Ge(n,γ)→71Ge ⟶EC71Ga. The activity of 71Ge in a NTD-Ge sample is obtained by measuring the event rates of the gallium KX-rays with energies of 9.2 and 10.3 keV using a Micro-Megas Detector (MMD) and a Silicon Drift Detector (SDD) individually, and incorporating with the correction of detection efficiencies estimated with the Monte Carlo simulation based on the GEANT4 toolkit. The measurements yield the total thermal neutron fluence with high precision, and the results are well consistent with each other between the MMD and SDD method.

Thermoluminescence properties of Dy3+-Doped Li2O–CaO–P2O5 glasses for neutron detection

In this study, we fabricated Dy3+-doped Li2O–CaO–P2O5 glasses using 6Li or nLi (nLi: containing 6Li or 7Li isotopes in natural abundance) to develop a new thermoluminescent material for neutron detection. Their thermoluminescence (TL) properties were investigated using X-ray and thermal neutron irradiation. TL glow peaks were observed at 366 K after X-ray irradiation, and the observed TL intensities of the nLi and 6Li glasses were similar, indicating that these materials have similar sensitivities to X-rays. A TL glow peak was observed at between 373 K and 400 K after neutron irradiation, and the TL intensities of the 6Li glass were significantly higher than those of the nLi glass, indicating that high-energy particles generated in the 6Li (n, α)3H reaction efficiently induced TL. The thermal neutron fluence and TL intensity derived from the 6Li (n, α)3H reaction in the neutron fluence range of 1.0 × 106–1.0 × 1011 neutrons/cm2 showed good linear correlation. These glasses are expected to be applied in TL dosimetry for X-rays, gamma rays, and thermal neutrons.

Estimating 10B and 14N doses using a pair of LiF-based TLD-600 and TLD-700 as an alternative to gold activation method

Purpose:A new method using TLD-600 and TLD-700 was proposed to measure a thermal neutron flux, which is used to evaluate 10B and 14N doses for boron neutron capture therapy (BNCT). Material and method:TLD-600 and TLD-700 were enriched by 6Li (95.6%) and 7Li (99.99%), respectively. The TLDs were calibrated using a 252Cf neutron source within a 30 cm-diameter D2O sphere to produce a thermal neutron flux calibration factor. The calibration factor was obtained by dividing the neutron-induced TLD readings by the delivered thermal neutron fluence. In addition, a neutron self-shielding correction was calculated to compensate for the perturbed neutron fluence between adjacent TLDs. To validate the procedure involved in this new method, TLDs were irradiated by thermal neutron fluxes from the research/educational reactor. The results were compared with the thermal neutron fluxes by the conventional gold activation method. Results and discussion:The calibration factor from TLDs irradiated by the thermal neutron flux from the 252Cf neutrons was determined 3.80 × 104 cm−2 nC−1. Using this factor, the thermal neuron flux at the reactor was measured 2.28 × 105 cm−2 W−1 s−1 by the TLD method. The one measured by the gold activation method was 2.17 × 105 cm−2 W−1 s−1 showing a difference of 5%. With the thermal neutron fluence measured by TLDs, a Li-6 dose of TLD-600 (Dn,αTLD) can be calculated. Then the two major components of BNCT dosimetry, a B-10 dose (D(n,α)B−10) and an N-14 dose (D(n,p)N−14) can be calculated by multiplying Dn,αTLD with constants due to a strong 1/v behavior of the three reactions. Conclusions:The TLD method developed in this study was successfully validated against an unknown mixed neutron/gamma field from the reactor, showing about 5% difference in measured thermal neutron fluences when compared to the conventional gold activation method. Therefore, we concluded that the two major dose components of BNCT, 10B and 14N doses can be accurately evaluated by the neutron-induced TLD dose.

Estimation with Monte Carlo of thermal neutrons in the FANT, using 252Cf and 241Am/9Be neutron source

The thermal neutron irradiation device (FANT), developed at the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid, is a high-density polyethylene regular parallelepiped, with a rather uniform neutron fluence inside its irradiation chamber. It uses a Am95241/Be49 neutron source aiming to provide thermal neutron fluence rates. Neutron spectra and neutron fluences were estimated with Monte Carlo methods in the FANT irradiation chamber when a Cf98252 neutron source is used and were compared with the results obtained with the Am95241/Be49 source. Regardless of the neutron source, the largest contribution is due to thermal neutrons, producing also epithermal and fast neutrons. Per neutron emitted by the source, the use of Cf98252 produces a larger amount of neutrons.

Reporting Dose in Boron Neutron Capture Therapy.

Reporting needs uniformity in definitions and terminology. This is crucial in order to be able to compare and aggregate clinical outcomes data across clinical trials. For decades, there is no standard for reporting dose in boron neutron capture therapy (BNCT). Multiple efforts were made for a common language reporting BNCT, mostly suggesting to report at relevant points the different dose components separately. Although this was accepted by some but not all clinicians, the situation remains an unsatisfactory and cumbersome reporting system, leading to confusion. Another suggestion is made here by proposing not to report the results of calculations that use a site-specific model with certain assumptions, but to report the input parameters needed for such calculations regardless of the model used. This will be mainly the thermal neutron fluence integrated over the irradiation time T and the average 10B concentration integrated over T.