Fluorescent Nuclear Track Detectors (FNTDs) provide high spatial resolution, wide linear energy transfer coverage, and reusability, making them well-suited for high-energy neutron dosimetry. When neutrons traverse a polyethylene converter, recoil protons are generated, and their tracks are stored inside the FNTDs and visualised through optical readout. Traditional analysis of FNTD images relies on deterministic algorithms or machine learning methods with explicit feature definition, limiting their general extension. In contrast, deep learning networks can extract image features enabling generalisation across different neutron energy spectra and dose values. In this study, a deep learning network was trained on images of FNTDs irradiated at six mono-energetic neutron energies and tested on images of FNTDs exposed to a broad-spectrum 241 Am-Be neutron source. Using raw images of irradiated FNTDs as input, the network predicted the proton tracks which were later counted. For the 241 Am-Be test dataset, a dose–response curve of identified tracks over ambient dose equivalent was fitted, and the sensitivity in terms of H ∗ ( 10 ) was extracted from the slope. When the fit was applied on the whole H ∗ ( 10 ) range, from 0 mSv up to 100 mSv, the predicted sensitivity for 241 Am-Be was S p r e d = ( 2280 ± 20 ) tracks mSv − 1 cm − 2 . The relative deviation of this predicted sensitivity from the reference sensitivity was 5.8%. When the fit considered only the H ∗ ( 10 ) range of the training dataset, namely from 5 mSv to 15 mSv, the predicted sensitivity for 241 Am-Be was S p r e d = ( 2500 ± 60 ) tracks mSv − 1 cm − 2 . This led to a relative deviation from the reference sensitivity of only 1.2%. Despite being trained solely on mono-energetic data, the model successfully generalised to the 241 Am-Be energy spectrum. • Fluorescent Nuclear Track Detector (FNTD) images segmented with deep learning. • The self-configuring nnU-Net framework was used for FNTDs. • FNTDs irradiated with mono-energetic neutrons were used as the training dataset. • Successful track identification of recoil protons. • FNTDs irradiated with broad spectrum 241 Am-Be properly predicted.

BackgroundSecondary neutrons are a major concern regarding side effects in ion beam therapy because they contribute to the out‐of‐field dose, particularly important for sensitive patient groups such as pregnant and pediatric patients. Measuring these neutrons is challenging because of their high kinetic energy, which is imparted to charged particles like fragments and recoil protons. In addition, accurate measurements require small detectors that ideally do not disturb the radiation field when measuring inside a phantom. Fluorescent Nuclear Track Detectors (FNTDs) have already shown promising results in ion beam dosimetry and the measurement in fast neutron fields. Given their high spatial resolution and sensitivity, FNTDs offer a promising approach for characterizing secondary neutron doses in complex radiation environments, such as those encountered in proton therapy.PurposeEstablish a methodology for estimating neutron‐induced out‐of‐field dose inside a phantom. The focus is to discuss the technical requirements and present initial experimental results from a proton treatment plan.MethodsThe analysis workflow for determining dose equivalent with FNTDs is introduced, including intensity‐to‐linear energy transfer (LET) in water conversion and track polar angle corrections. FNTDs were placed inside RW3 and polymethyl methacrylate phantoms and irradiated with a proton spread‐out Bragg peak (SOBP) plan. Experimental results from two downstream positions in each phantom were used to benchmark Monte Carlo simulations.ResultsA polar angle correction function was established, indicating intensity corrections of approximately a factor of 2 at and up to a factor of 3.5 beyond . Furthermore, a few short‐range high‐LET tracks with a low probability of occurrence have been found. Despite accounting for only about 1% of the total fluence, high‐LET tracks can contribute more than 50% of the total dose equivalent. When not considering these short‐range tracks, the relative agreement in dose equivalent between simulations and experiments was within (1.10±0.10) to (1.49±0.13).ConclusionsThis work presents the first LET‐based method using FNTDs to estimate out‐of‐field neutron dose for a proton SOBP plan, measured inside a phantom. Integrating this method into clinical workflows may improve out‐of‐field dose estimation for sensitive patient groups, such as pregnant or pediatric patients, by enabling prior dose assessments using anthropomorphic phantoms.

This study presents a comparative analysis between Monte Carlo simulations (MCNP-5) and experimental measurements of thermal neutron fluence using gold foil activation at the TNF2 irradiation facility of the Neutron Metrology Laboratory (LNMRI/IRD). The MCNP-5 model was constructed to replicate the irradiator geometry and source configuration. Gold foils, both bare and cadmium-covered, were irradiated and analyzed via gamma spectrometry to determine the induced activity of 198Au. The results of the calculated nuclear reaction rates were compared to the measured activities through C/E (Calculated/Experimental) analysis. The findings show consistent agreement within experimental uncertainty, supporting the accuracy of the MCNP model and reinforcing the role of TNF2 in the national and international metrological infrastructure for neutron dosimetry. Future developments will include improvements in flux mapping and personal dosimetry applications.

Comprehensive review of neutron techniques, detection, and dosimetry in science and technology

Neutron and gamma dosimetry with radiochromic film

Advanced Monte Carlo analysis of ex-vessel neutron dosimetry for cycle 25 of Krško nuclear power plant

This study aims to measure albedo neutron doses arising during high-energy photon radiotherapy, using an albedo TLD to check the spatial distribution and dose-dependency of these neutrons across a tissue-equivalent phantom. Irradiations were performed with a 15 MV medical linear accelerator. Alnor-type albedo TLD badges containing MTS-6 (6LiF, neutron and photon sensitive) and MTS-7 (7LiF, only photon sensitive) which were calibrated before Linac irradiation of an Alderson Rando phantom placing the badges at various anatomical sites. The design allows selective measurement of albedo neutrons by a dedicated window exposing a pair of MTS - 6 & 7 crystals, while others remain shielded by boron-loaded plastic. Dose measurements were performed at the lung center, right and left lungs, head-neck, and upper abdomen for delivered photon doses of 100, 200, and 300 cGy. Reflected neutron dose was highest at the beam center (lung center: 440.9 – 520.4 mSv) and decreased steeply at lateral (right/left lung: 33.0 – 66.8 mSv) and peripheral (head-neck, upper abdomen: 4.2 – 13.3 mSv) positions. At all locations, neutron dose increased linearly with delivered photon dose (R² ≈ 1). The dose at the peripheral regions confirms rapid spatial attenuation. The use of albedo TLD badge proves its effectiveness for distinguishing reflected neutrons from direct field and photon doses. Albedo neutrons contribute a measurable amount to total delivered dose particularly within and near the field. Routine assessment of albedo neutron dosimetry is recommended to ensure treatment quality. Ban. J. Med. Phys., Vol -16, Issue -1, 2025 : 7-10

Objective.Isochronous cyclotrons, synchrocyclotrons, and synchrotrons are used to accelerate protons for proton therapy. An accurate measurement of neutron doses generated by these accelerators and associated delivery systems and its clinical relevance requires systematic protocols and proper neutron dosimetry for a meaningful assessment. We present the first comprehensive comparison of neutron ambient dose equivalent (H*(10)) produced by clinically operational proton therapy systems.Approach.Treatment plans with 10 cm modulation-depth and ranges of 10 cm (R10M10) and 25 cm (R25M10) were created to cover a 10 × 10 × 10 cm3water target. The pencil beam scanning proton therapy machines studied were: two gantry-mounted synchrocyclotrons (Hyperscan, Mevion, half-gantry), two isochronous cyclotrons (ProBeam, Varian, full-gantry), one isochronous cyclotron (Proteus, IBA, full-gantry), and two synchrotrons (PROBEAT, Hitachi, full- and half-gantry). Proton beams were delivered to 30 × 30 × 40 cm3plastic water phantoms. WENDI-II and LUPIN-BF3-NP neutron rem-meters were positioned at three angles (0°, 45°, 90°) relative to the beam direction to measure the neutron H*(10) at distances between 50-300 cm from the isocenter.Main results.H*(10) showed dependence on beam energy, machine type, and measurement location. The highest reading was for the gantry-mounted synchrocyclotron, whereas other systems produced approximately comparable neutron doses. In all cases, the H*(10) reduced with distance from the isocenter. The H*(10) drop at 2 m distance compared to that at 0.5 m was a factor of ∼5 for the gantry-mounted synchrocyclotron whereas in other systems the decrease was a factor of 10. The WENDI-II device suffered from dead-time-associated under-estimation of the dose by a factor of ∼2-3 under the synchrocyclotron beam due to its high dose-per-pulse. However, WENDI-II and LUPIN-BF3-NP results were within reasonable agreement in isochronous cyclotron and synchrotron beams, indicating that both devices are suitable for those systems.Significance.Neutron H*(10) is dependent on various parameters including beam energy, measurement location, as well as machine design. Caution must be exercised in choosing the appropriate neutron-dose-measurement device to be used for low-duty-factor, particularly in high-instantaneous-rate proton delivery systems. By delivering the same volumetric proton dose across different machines, this work provides a benchmark for inter-system comparisons and serves as a foundation for future studies.

This study investigates neutron radiation sources in medical cyclotrons used for PET isotope production, focusing on differences between 18F and 11C. Neutron and gamma dose rates were measured in the bunker and operator control room during routine production with an 11 MeV Eclipse cyclotron. 18F production generated approximately 2.5 times higher neutron levels in the bunker than 11C. Shielding performance also varied: the same wall reduced neutron fluxes by factors of kF = 14,000 for 18F and kC = 86,000 for 11C, while gamma shielding was similar for both isotopes (kγ ≈ 28,000). However, the neutron shielding factor calculated from the data for 18F should be taken as kF ≥ 1.4 × 104, because several neutron readings reached the upper limit of the detector range, which indicates a partial underestimation of the dose in the bunker. Consequently, neutron levels in the control room during 18F production were about 15-fold higher than during 11C production. These differences result from distinct neutron generation mechanisms. The 18O(p,n)18F reaction produces primary neutrons with a Maxwellian spectrum (~2.5 MeV), while 11C neutrons arise solely from secondary interactions in structural materials. The findings emphasize the need for composite shielding adapted to isotope-specific spectra. Annual dose estimates (260 18F and 52 11C productions) showed neutron exposure (3.78 mSv/year, 57%) exceeded gamma exposure (2.82 mSv/year, 43%). The total dose of 6.6 mSv/year is ~33% of regulatory limits, supporting compliance but underscoring the need for dedicated neutron dosimetry.

This study aims to evaluate the potential of polyallyl diglycol carbonate (PADC) detectors, coupled with 6 LiF converters, for passive environmental neutron dosimetry. Their performance is compared against three other methods, fission track detectors, thermoluminescent detectors (TLDs), and optically stimulated luminescence detectors (OSLDs), to determine their viability for passive routine neutron monitoring in environmental settings. The four detector types investigated (PADCs, fission tracks, TLDs and OSLDs) were placed in a moderator sphere and exposed to an AmBe neutron field. The measurements evaluated the dose response, measurement range, and precision of each system. Results confirm that the advantage of the track detectors (PADC or fission track) is their gamma insensitivity: the TLD and OSLD methods require the combination of neutron-sensitive with neutron-insensitive detectors, as both are gamma-sensitive. This translated into a detection limit H ∗ ( 10 ) =6 μ Sv for PADC detectors coupled with 6 LiF converters, comparable to that of fission track detectors (3 μ Sv). In contrast, the gamma-sensitive luminescence systems showed a detection limit of 50 μ Sv for the TLDs and 180 μ Sv for the OSLDs under AmBe neutron irradiations. This worsens under simulated environmental conditions, where a significant gamma contribution is expected, as demonstrated by a test where mixed gamma/neutron irradiation was used. These results indicate the potential of the PADC coupled with a 6 LiF converter system as a robust alternative to fission track detectors for passive environmental neutron dosimetry. • A novel environmental neutron dosimetry system using PADCs coupled with 6 LiF converters was tested. • Performance was benchmarked against fission track, TLD, and OSLD systems in AmBe neutron fields. • PADCs and fission tracks showed the lowest detection limits ( H *(10) = 6 μ Sv and 3 μ Sv, respectively). • PADCs and fission track systems are insensitive to gamma, improving detection in mixed fields.

Accurate neutron detection in mixed photon-neutron and pulsed radiation fields is technically challenging, impacting industrial and medical applications. This paper presents the first measurements of thermal neutrons in conventional radiotherapy accelerators using a silicon carbide (SiC) P–N diode with different neutron converters. SiC detectors enable real-time estimation of secondary thermal neutron contributions, crucial for emerging radiotherapy techniques requiring precise neutron fluence monitoring. Beyond medical applications, the presented detectors show potential for neutron dosimetry, radiation monitoring, nuclear safety, and scientific research. The SiC diode active detection layer is less than 30 µm thick, and provides excellent gamma rejection (5times 10^{-8}), allowing discrimination of neutrons-induced events in mixed radiation fields. Experimental tests conducted on a TrueBeam radiotherapy LINAC demonstrated a thermal neutron detection efficiency of (4.32 ± 0.02)% for a (50 ± 10) µm thick ^6LiF neutron converter. The detector, placed at 1.2 m from the accelerator isocenter, was used to measure neutron fluences at different monitor unit (MU) rates, ranging from 100 to 600 MU/min, with the LINAC operating at 15 MV. Under these conditions, the detector exhibited good linearity, without saturation or dead time effects.

Advancements in neutron dosimetry: A comprehensive FLUKA simulation of PLASTIC scintillator response across multiple irradiation geometries.

Commissioning of the H2 neutron irradiation facility at MARIA research reactor, as a part of MARIA Neutron Laboratory.

The response of the borated CR-39 detector (a registered trademark of TASTRACK Industries) has been investigated with the aim of studying its application in both fast and thermal neutron dosimetry. Optimization tests on the neutron sensitivity and background tracks were performed to ensure the detector’s acceptable sensitivity to the personal dose equivalent of fast and thermal neutrons. The detectors were irradiated to dose equivalents of reference neutron sources, with a 241Am-Be source for fast neutrons and a miniature neutron source reactor for thermal neutrons. The irradiated detectors were processed with various chemical etching parameters to optimize their performance to an acceptable level. The chemical etching procedures were carried out for 4 h and 6 h with a KOH water solution and for 6 h and 9 h with a NaOH water solution at 70°C for all detector samples. The results showed good repeatability and remarkable signal-to-noise ratio values were achieved in etching for 6 h and 9 h with the NaOH water solution of 7 M at 70°C. By selecting these parameters for chemical etching, the detector’s efficiency increased by approximately a factor of 2. Thus, the borated CR-39 material maintained reliable performance for both fast/thermal neutrons through the same optimized chemical etching parameters, 6 h with 7-M NaOH water solutions at 70°C. Through this study, we achieved an excellent response from the borated CR-39 to thermal neutrons. We improved the detector’s efficiency by shortening the etching time without enhancing the etching temperature or etchant concentration. These optimized parameters demonstrated consistent results, ensuring the detector’s ability to measure equivalent neutron doses across different environments without needing specific adjustments to the etching conditions.

Bhabha Atomic Research Centre participated in the EURADOS Inter-comparison Exercises IC2022ph [Hp(10) and Hp(0.07)] and IC2022n [Hp(10] for assessment of external exposures, using the CaSO4:Dy-based thermoluminescent dosemeter (TLD) (indigenously developed) for photon and beta radiations and etched track neutron dosemeter, respectively. The performance of TLD personnel dosimetry system, assessed in accordance with ISO 14146:2018 and ANSI N13-11.2009 standards, is satisfactory. The results of the IC2022ph have enhanced confidence in the accuracy and reliability of the Hp(10) and Hp(0.07) algorithms used for TLD badge. Dose evaluation for the neutron dosimetry system was based on in-house developed imaging system. The performance of neutron dosemeter system, as assessed in this interlaboratory comparison, indicates a need to improve the quality of the detector material. Participation in the inter-comparison exercises underscores the reliability of dosimetry systems and highlights their effectiveness in accurately assessing external radiation exposures in India's personnel monitoring programs.

Ionizing Radiation Metrology Laboratory (IRML), National Institute of Standards (NIS), Egypt is developing its capabilities in neutron metrology to the primary standard level. Recently, a Mn-bath system was installed at IRML. MCNP6 was used to simulate physical processes, interactions, and calculations of the main parameters used in primary standard neutron dosimetry. MCNP6 was used to calculate Am-Be neutron flux considering its constitution mixture and capsule dimensions and neutron doses at different depths within the Mn bath. The effect of geometry, and composite material for various ratios of hydrogen to Manganese atoms (NH/NMn) within the solution of MnSo4 were considered in the simulation, in addition to the determination of neutron energy reduction within the Mn-bath. The variance of thermalized neutron energy due to their interaction with water and capture by MnSO4 solution was calculated. The induced gamma radiation from the 55 Mn (n, γ) 56 Mn interaction was calculated. It was found that neutron doses calculated using MCNP6 are incompatible with those measured by the standard neutron monitors. The average energy distribution decreases as depth increases. The induced gamma activity calculated from neutron interaction within the Mn bath agrees with the experimental results and literature. • MCNP6 was used to calculate neutron flux delivered by Americium-Beryllium (Am-Be) source. • MCNP6 was used to calculate main parameters used in primary standard neutron dosimetry. • MCNP6 was used to calculate neutron flux from the neutron source, neutron interactions with materials in the Mn-bath. • MCNP6 was used to calculate background radiation around the Mn bath and also the induced gamma activity.

Feasibility study for establishing a linear and wide range neutron dosimetry field based on MNSR neutron radiography beamline

In ion beam radiotherapy, treatment radiation fields are inevitably contaminated with secondary neutrons. The energies of these neutrons can reach several hundreds of MeV. Fluorescent nuclear track detectors (FNTDs) offer a promising solution for dosimetry of fast and high-energy neutrons, particularly given their low linear energy transfer in water (LET) detectionthreshold. This study presents an experimental FNTD sensitivity analysis in six fast mono-energetic neutron fields, comparing the response to poly allyl diglycol carbonate (PADC) neutron detectors, and investigates the feasibility of estimating ambient dose equivalent for neutrons, (10). Moreover, it investigates the impact of converter thickness on the detector signal for both fast and high-energy neutrons and analyzes the resulting differences insignal. FNTDs and PADCs were exposed to mono-energetic neutron fields with energies of 1.2 MeV, 2.5 MeV, 5 MeV, 6.5 MeV, 14.8 MeV, and 19MeV and evaluated based on the track density. The (10) values for FNTDs were determined by applying energy calibration factors, , which were determined through Monte Carlo (MC) simulations. The benchmarked MC model is employed to investigate the sensitivity of FNTDs to high-energy neutrons up to 200MeV for various polyethylene (PE) converter thicknesses and to analyze the detector signal, including the particle type and the recoil protonLET. The sensitivity values revealed an energy dependence for FNTDs, with variations by a factor of up to 23, whereas PADC detectors showed a smaller variation, ranging from 3 to 12. Accurate (10) estimation can be achieved employing MC-derived factors, with deviations not exceeding . The sensitivity values increased almost continuously up to for PE converter thicknesses above , whereas plateaued for thinner PE converters above 10 MeV to 15MeV. For neutrons above , the generated fragments are deuterons, tritons and , which constitutes up to or more of the total fluence in a neutron field. The recoil proton LET dropped from approximately to nearly one order of magnitude less between 1.2 MeV and 19MeV, with an average LET of approximately at . This study compares FNTD and PADC detector sensitivities, demonstrating a notable energy and converter thickness dependence for FNTDs, which is essential for precise dosimetry. Accurate (10) values for fast mono-energetic neutrons up to were determined utilizing MC simulations. A benchmarked MC model for fast neutrons was then applied to analyze the FNTD signal for high-energyneutrons.

H*10 neutron dosimetry (unlike gamma dosimetry), requires consideration of neutron energy spectra due to the 20× variation of the weight factor over the thermal-to-fast energy range, as well as the neutron radiation field dose rates ranging from cosmic, ~.01 μSv h -1 levels to commonly encountered ~10-200 μSv h -1 in nuclear laboratories/processing plants, and upwards of 10 4 Sv h -1 in nuclear reactor environments. This paper discusses the outcome of the comparison of spectrum-weighted neutron dosimetry covering thermal-to-fast energy using the novel H*-TMFD spectroscopy-enabled sensor system in comparison with measurements using state-of-the-art neutron dosimetry systems at SRNS-Rotating Spectrometer (ROSPEC), and non-spectroscopic Eberline ASP2E ("Eberline") and Ludlum 42-49B ("Ludlum") survey instrumentation. The H*-TMFD was validated for gamma blindness using a 2.48×10 10 Bq 137 Cs source. The background dose rate in Savannah River Nuclear Solutions' (SRNS) low-scatter facility with all neutron sources withdrawn was estimated at 0.005 μSv h -1 . From moderately high radiation field tests conducted with the high intensity (1.4 × 10 9 n s -1 ) 252 Cf source and a total data collection time of ~0.15 h, the predicted dose rates from Eberline (non-spectroscopic), Ludlum (non-spectroscopic), and spectroscopic H*-TMFD instruments were found to be: ~170 μSv h -1 , ~200 μSv h - , and ~ 120 μSv h -1 , respectively. The equivalent spectroscopic (SRNS measured) H*10 dose rate from ROSPEC value is 130 μSv h -1 , within 10% of H*10-TMFD measurement. Tests conducted for ultra-low intensity radiation field used a ~ 1.6 × 10 3 n s -1 252 Cf bare neutron source for which over a collection time of ~18 h, the Eberline meter measured an instantaneous dose/count rate of 0 μSv h -1 and a pulse-integrated dose rate of 0.034 μSv h -1 at ~1 m. In contrast, the H*-TMFD panel located 0.22 m in direct line of sight of the 252 Cf source spectroscopically measured ~0.4 μSv h -1 (within +/- 5%) over 1.8 h collection live time-with which spectrum matched perfectly to that of a bare 252 Cf source. The H*TMFD predicted value of ~0.4 μSv h -1 was cross-checked and found to be within 10% of LLNL's published value of ~0.37 μSv h -1 (intensity/distance corrected via 1/r 2 law of 25.5 μSv h -1 at 1 m for a 1 μg 252 Cf source); as well as from use of ICRP 74 conversion coefficients and MCNP code simulations. As expected, for a bare 252 Cf source, H*TMFD measured epithermal neutron energy-related dose rates are well below 1% of the total dose rates. For ~0.01 μSv h -1 neutron radiation fields, ROSPEC measurements for H*10 dose rates are estimated to take 7+ d, vs. under 2 h with the H*TMFD. The feasibility of using a single CTMFD in survey mode for H*10 dose rate (nSv h - to μSv h -1 ) measurements within 2-3 min is demonstrated.

The CORANI (KAERI/CEA COllaboration for Research reactor Application of Neutron dosimetry and Instrumentations) project is a collaborative effort between KAERI and CEA aimed at validating advanced neutron instrumentation and dosimetry techniques at the High-Flux Advanced Neutron Application Reactor (HANARO). The Phase 1 campaign focuses on neutron field characterization using self-powered neutron detectors and fission chambers, complemented by activation-foil dosimetry. This paper presents the feasibility study results for CORANI Phase 1 to confirm the reliability of the adopted methodologies and provide a solid basis for experimental conditions and irradiation rigs. The preliminary results demonstrate well agreement between the two different measurement systems and confirm the operational parameters for the forthcoming Phase 1 campaign scheduled for the fourth quarter of 2025.