Neutron Dose Components Research Articles

Measurements of the neutron and gamma dose via TL/OSL in LiF:Mg,Ti and activation methods analysis of aluminum foils

The measurement of beta/gamma and neutron dose components in mixed radiation fields is described involving combined OSL and TL measurements. The accuracy of the technique is investigated by irradiation of LiF:Mg,Ti samples with 14.1 MeV neutrons and a comparison of the OSL/TL results with the neutron fluence determined by Al activation. The precision of the measurements is 6.1 to 6.7 % (1 SD) for the OSL/TL technique compared to 7.3 % for the Al foils. The advantage of the technique is the simultaneous determination of the two dose components on the same samples of 7LiF:Mg,Ti in a single dedicated TL/OSL reader.

Comparison of biological weighting functions to estimate the microdosimetric RBE in BNCT.

In the framework of the MUNES project, a prototype accelerator-based thermal neutron source was developed and installed at the Legnaro National Laboratories of INFN. The microdosimetric characterization of this radiation field was performed with a Tissue-Equivalent Proportional Counter with interchangeable cathode walls, either doped with 100ppm of 10B or without boron doping. A suitable subtraction procedure allowed to discriminate the gamma, neutron and BNC dose components (Selva etal., 2022, Appl. Radiat. Isot. 182, 110144). The measured microdosimetric spectra can be weighted with a biological weighting function to estimate the Relative Biological Effectiveness of the radiation field, for the purpose of intercomparison between different thermal neutron sources. This work compares, therefore, the biological doses resulting from four different weighting functions applied to the same initial microdosimetric spectrum, discussing strengths and limitations of each of them.

A Novel, Rapid Response Renewable Biopolymer Neutron and Gamma Radiation Solid-State Detector for Dosimetry and Nuclear Reactor Flux-Power Mapping

A novel solid-state neutron and gamma radiation monitor-dosimeter based on biopolymer polylactic acid (PLA) is presented. The resulting detector (PLAD) technology takes advantage of property changes of the renewable PLA resin when subject to ionizing nuclear radiation. A simple yet rapid and accurate (±10%) low-cost (<$0.01/detector) mass loss upon dissolution (MLD) technique was successfully developed; MLD is based on a simple mass balance for discerning neutron and/or gamma doses using small (40 mg, ~4 mm diameter) ultra-low-cost (<$0.01) resin beads via dissolution in acetone. The GammaCellTM Co-60 irradiator, and the PUR-1 12 kW fission nuclear research reactor were utilized, respectively. Irradiation absorbed doses ranged from 1 to 100 kGy. Acetone bath temperature was varied from ~40 °C to ~54 °C. Results revealed a strong dependence of MLD on acetone bath temperature between neutron and gamma photon dose components; this allowed for the unique ability of PLAD to potentially perform as both a neutron-cum-gamma or as a gamma or neutron radiation dosimeter and intensity level detector. A linear trend is found for combined neutron and gamma radiation doses from 0 to 40 kGy when dissolution is conducted above 50 °C. The important potential ability to distinguish neutron from gamma radiation fields was scoped and found to be feasible by determining MLD at 45 °C. The potential was studied for simultaneous use as an in-core neutron and gamma monitor of an operating 3 GWt light-water reactor (LWR). Scoping tests were conducted with the pre-irradiated (@ 20 °C) PLAD resin beads followed by heating to in-core LWR coolant (300 °C) conditions for ~30 s corresponding to the time to reach ~40 kGy total doses in a typical 3 GWt LWR. MLD results were unaffected, indicating the exciting and unique potential for in situ (low-cost, accurate and rapid) simultaneous mapping of neutron and gamma radiation fluxes, related dosimetry, and fission power level monitoring.

Study of thermally and optically stimulated luminescence in LiF:Mg,Ti following neutron and beta irradiation

The results reported herein were motivated by the potential application of combined optically stimulated (OSL) and thermoluminescence (TL) measurements for separation between the beta and neutron dose components in mixed neutron-beta radiation fields. The advantages of OSL/TL are two-fold: (i) The OSL and TL readout can be carried out on the same sample and (ii) the greater efficiency of OSL to high ionization density (HID) radiation due to the excitation of F2 and F3 + centers. The beta calibration coefficients for LiF:Mg,Ti (TLD-700) were measured using a 90Sr/90Y source calibrated at the Soreq Nuclear Research Center nuclear facility. The estimation of the neutron calibration coefficients was carried out by irradiation with broad beam fast neutrons of median energy 5 MeV and a quasi-monoenergetic neutron field of mean energy 14.8 MeV at the Physikalisch Technische Bundenanstalt. The preferential excitation of OSL following HID irradiation has been demonstrated.

Clinical 6 MV X-Ray facility photo-neutron/fission interrogations with TMFD sensors

Detection of several targets (U, Be, D2O) of interest to national security has been achieved at Purdue University’s 6 MV X-ray clinical linear accelerator (CLINAC), from photoneutron/fission induced neutron signatures using a Centrifugally Tensioned Metastable Fluid Detector (CTMFD). Traditional neutron detection schemes were saturated and ineffective under the intense (to ∼104 R/h) photon environment; however, the CTMFD sensor technology was capable of measuring the relatively low (∼1012x lower intensity) neutron field produced from 6 MeV-end point X-ray beam interactions with the CLINAC components and surroundings. This is in stark contrast to published results based on the conventional belief that the neutron dose component in 6 MV CLINAC facilities is immeasurable over the background photon background dose. Photoneutron and photofission neutron production rates were detected and analyzed at various standoffs from the isocenter, with and without special nuclear material (SNM) related targets ranging in mass from 5 g of Be, to ∼0.6 kg of UO2. It was confirmed that 6 MV X-ray photons do result in measurable neutron dose above normal cosmic background. Additionally, in threshold mode, the CTMFD operating under a tensioned metastable pressure of -5 bar, enabled conclusive rejection of background photoneutrons, while enabling the rapid detection of: Be, D2O, C and U targets within seconds to minutes while operating in and around 104 R/h photon fields.

Japanese pediatric and adult atomic bomb survivor dosimetry: potential improvements using the J45 phantom series and modern Monte Carlo transport

The radiation exposure estimates for the atomic bomb survivors at Hiroshima and Nagasaki have evolved over the past several decades, reflecting a constant strive by the Radiation Effects Research Foundation (RERF) to provide thorough dosimetry to their cohort. Recently, a working group has introduced a new series of anatomical models, called the J45 phantom series, which improves upon those currently used at RERF through greater age resolution, sex distinction, anatomical realism, and organ dose availability. To evaluate the potential dosimetry improvements that would arise from their use in an RERF Dosimetry System, organ doses in the J45 series are evaluated here using environmental fluence data for 20 generalized survivor scenarios pulled directly from the current dosimetry system. The energy- and angle-dependent gamma and neutron fluences were converted to a source term for use in MCNP6, a modern Monte Carlo radiation transport code. Overall, the updated phantom series would be expected to provide dose improvements to several important organs, including the active marrow, colon, and stomach wall (up to 20, 20, and 15% impact on total dose, respectively). The impacts were especially significant for neutron dose estimates (up to a two-fold difference) and within organs which were unavailable in the previous phantom series. These impacts were consistent across the 20 scenarios and are potentially even greater when biological effectiveness of the neutron dose component is considered. The entirety of the dosimetry results for all organs are available as supplementary data, providing confident justification for potential future DS workflows utilizing the J45 phantom series.

Numerical analysis of the alanine response using Monte Carlo: Correlation with experimental results

International centers based on research reactors (ICERR) project supported by IAEA, offers training and research in the application of advanced nuclear reactor technology. In this context of the project, alanine dosimeters were irradiated under mixed neutron/gamma irradiation in the core of the MINERVE research reactor at the CEA Cadarache. The integrated dose of the alanine was measured by electron spin resonance spectrometry with reference to an absorbed dose in water from a 60Co gamma-ray beam. Neutron and gamma dose components were calculated using the MCNP code. The neutron dose and relative effectiveness (REn) for neutrons was evaluated using FLUKA code and Track Structure Theory (TST) formulations of alanine response towards irradiations with ions produced by neutron interactions with alanine. REn values were found to be between 0.516 and 0.538. The measured dose was compared to the calculation results. The best agreement between the alanine integral measurements and model predictions (0.4%) was obtained by using the Cucinotta alanine detector response model.

Machine learning methodology for high throughput personalized neutron dose reconstruction in mixed neutron\u2009+\u2009photon exposures

We implemented machine learning in the radiation biodosimetry field to quantitatively reconstruct neutron doses in mixed neutron + photon exposures, which are expected in improvised nuclear device detonations. Such individualized reconstructions are crucial for triage and treatment because neutrons are more biologically damaging than photons. We used a high-throughput micronucleus assay with automated scanning/imaging on lymphocytes from human blood ex-vivo irradiated with 44 different combinations of 0–4 Gy neutrons and 0–15 Gy photons (542 blood samples), which include reanalysis of past experiments. We developed several metrics that describe micronuclei/cell probability distributions in binucleated cells, and used them as predictors in random forest (RF) and XGboost machine learning analyses to reconstruct the neutron dose in each sample. The probability of “overfitting” was minimized by training both algorithms with repeated cross-validation on a randomly-selected subset of the data, and measuring performance on the rest. RF achieved the best performance. Mean R2 for actual vs. reconstructed neutron doses over 300 random training/testing splits was 0.869 (range 0.761 to 0.919) and root mean squared error was 0.239 (0.195 to 0.351) Gy. These results demonstrate the promising potential of machine learning to reconstruct the neutron dose component in clinically-relevant complex radiation exposure scenarios.

In-phantom neutron dose measurement using Gafchromic film dosimeter for QA of BNCT beams

In order to improve the quality assurance (QA) procedure for beams of boron neutron capture therapy (BNCT), this study introduced using the Gafchromic film dosimeter for neutron dose measurement of BNCT beams. The crucial part of this study was investigating an approach to employ the Gafchromic film dosimeter placed inside a PMMA phantom irradiated by a BNCT beam. The spatial distribution of neutron dose of the film was determined using measurements and Monte Carlo calculations. By employing the present approach, the two-dimensional distributions of the neutron dose component of the film at specific depths in the phantom were successfully obtained. The determined neutron dose profiles were in good agreement with the calculated ones. This study also confirmed the finding that the film dosimeter is sensitive to thermal neutrons by comparing the depth-capture-reaction-rate and depth-dose distributions. Results of this work not only proved the feasibility of using the proposed method for the QA measurement of beam delivery but also revealed the advantages of easy-handling and remarkable spatial resolution of the film dosimeter when applied to BNCT fields. The present work can help to verify the dose uniformity and output stability of BNCT beams prior to clinical irradiation.

Geant4 beam model for boron neutron capture therapy: investigation of neutron dose components.

Boron neutron capture therapy (BNCT) is a biochemically-targeted type of radiotherapy, selectively delivering localized dose to tumour cells diffused in normal tissue, while minimizing normal tissue toxicity. BNCT is based on thermal neutron capture by stable [Formula: see text]B nuclei resulting in emission of short-ranged alpha particles and recoil [Formula: see text]Li nuclei. The purpose of the current work was to develop and validate a Monte Carlo BNCT beam model and to investigate contribution of individual dose components resulting of neutron interactions. A neutron beam model was developed in Geant4 and validated against published data. The neutron beam spectrum, obtained from literature for a cyclotron-produced beam, was irradiated to a water phantom with boron concentrations of 100 μg/g. The calculated percentage depth dose curves (PDDs) in the phantom were compared with published data to validate the beam model in terms of total and boron depth dose deposition. Subsequently, two sensitivity studies were conducted to quantify the impact of: (1) neutron beam spectrum, and (2) various boron concentrations on the boron dose component. Good agreement was achieved between the calculated and measured neutron beam PDDs (within 1%). The resulting boron depth dose deposition was also in agreement with measured data. The sensitivity study of several boron concentrations showed that the calculated boron dose gradually converged beyond 100 μg/g boron concentration. This results suggest that 100μg/g tumour boron concentration may be optimal and above this value limited increase in boron dose is expected for a given neutron flux.

Radiobiological response of U251MG, CHO-K1 and V79 cell lines to accelerator-based boron neutron capture therapy.

In the current article, we provide in vitro efficacy evaluation of a unique accelerator-based neutron source, constructed at the Budker Institute of Nuclear Physics (Novosibirsk, Russian Federation), for boron neutron capture therapy (BNCT), which is particularly effective in the case of invasive cancers. U251MG, CHO-K1 and V79 cells were incubated and irradiated in various concentrations of boric acid with epithermal neutrons for 2–3 h in a plexiglass phantom, using 2.0 MeV proton energy and 1.5–3.0 mA proton current, resulting in a neutron fluence of 2.16 × 1012 cm−2. The survival curves of cells loaded with boron were normalized to those irradiated without boron (to exclude the influence of the fast neutron and gamma dose components) and fit to the linear–quadratic (LQ) model. Colony formation assays showed the following cell survival rates (means ± SDs): CHO-K1: 0.348 ± 0.069 (10 ppm), 0.058 ± 0.017 (20 ppm), 0.018 ± 0.005 (40 ppm); V79: 0.476 ± 0.160 (10 ppm), 0.346 ± 0.053 (20 ppm), 0.078 ± 0.015 (40 ppm); and U251MG: 0.311 ± 0.061 (10 ppm), 0.131 ± 0.022 (20 ppm), 0.020 ± 0.010 (40 ppm). The difference between treated cells and controls was significant in all cases (P < 0.01) and confirmed that the neutron source and irradiation regimen were sufficient for control over cell colony formation. We believe our study will serve as a model for ongoing in vitro experiments on neutron capture therapy to advance in this area for further development of accelerator-based BNCT into the clinical phase.

Dose reassessment of LiF:Mg,Ti detectors in the mixed fields

Abstract Phototransfered thermoluminescence (PTTL) found its use in radiation dosimetry for dose reassessment primarily in photon radiation fields. After the dose has first been evaluated by a “normal” thermoluminescence (TL) measurement, the same detector is subjected to the PTTL dose reassessment procedure providing a new estimation by an independent measurement procedure. PTTL also gives a new possibility for a dose reassessment and for separate dose determination in the mixed fields. It was shown in our previous papers that it is possible to determine separately gamma and neutron dose components with a single detector due to different sensitivities of TL and subsequent PTTL to neutrons in comparison to gamma rays. The main disadvantage, the low sensitivity of both, TL and PTTL, to neutrons restricts the use only for the case of radiation accident. In this work we have investigated the possibilities of improvement the sensitivities of PTTL by heating the samples during UV irradiation and by using the sensitization technique involving a “pre-dose” followed by annealing. Two types of LiF:Mg,Ti detectors, TLD-100 and TLD-700, were examined by irradiation in 137 Cs gamma ray field and with neutrons of 14.5 MeV. After UV irradiation at different temperatures, PTTL increases reaching the values 2.1 and 2.5 times higher at 90 °C than that at 25 °C for TLD-100 and TLD-700 respectively. Increasing level is approximately the same for gamma rays and neutrons. After sensitization, TL signal is one half of that before sensitization. PTTL signal has increased 11 times for gamma rays and 5 times for neutrons at 25 °C with significant increase with increasing temperature during UV irradiation.

SU‐FF‐T‐61: Determination of Neutron and Gamma Dose Rates in Water Surrounding a New Interstitial Cf‐252 Brachytherapy Source

Purpose: A new generation of high‐activity miniature 252Cf sources for interstitial brachytherapy has recently been encapsulated at the Oak Ridge National Laboratory (ORNL). The purpose of this study is to determine the neutron and gamma dose rates in water surrounding the new 252Cf source. Method &amp; Materials: The neutron and gamma dose components of the new 252Cf source in water were determined by the measurements of two miniature ion chambers (T1 and M1 of Standard Imaging, Inc.) that have similar gamma responses but very different neutron responses. A Lucite‐walled water phantom was built to conduct the experiment. The experimental setup was also modeled by the Monte Carlo transport code MCNP. Results: The results show that the measured neutron absorbed dose rates were approximately 25% lower than that predicted by MCNP for all dose positions in water. The measured gamma absorbed dose rates in water, on the contrary, are higher than that predicted by MCNP. The differences between the measured and MCNP‐predicted gamma dose rates are most pronounced (∼a factor of two) at the distance of 1 cm, and fall to approximately 30% at distances 2 cm and beyond. Conclusions: The measured neutron and gamma dose rates do not agree with the dose rates predicted by MCNP. The discrepancies between the two results suggest that the true 252Cf content of the new source is approximately 25% less than the ORNL estimate and that the new source emits significantly more low‐energy gamma rays than the large 252Cf sources (e.g. the AT source) used in the previous studies. Acknowledgement: This research is supported by Isotron Inc.

A virtual model of the patient's head for BNCT

A virtual model of the patient's head for BNCT The aim of the present work was creating a virtual phantom of a human head for BNCT, as a part of the BNCT programme project. This model is an amplification of the simple model described in earlier publications. It takes into account the major head organs as well as the scalp and skull. The chemical composition of all tissues was modelled according to the recommendations of the ICRP. The organs were parameterized using mathematical formulas based on the human head magnetic resonance images. The model was used for calculating the thermal neutron flux and the injuring (fast neutron, nitrogen and gamma) dose components for the head irradiated using the therapeutic neutron beam, whose parameters were obtained as the result of the modelling of the filter/moderator system for the BNCT therapeutic beam from the MARIA reactor.

TL and PTTL of TLD-100 and TLD-700 after irradiation with Pu–Be neutrons

Phototransfered thermoluminescence (PTTL) found its use in radiation dosimetry for dose reassessment. After the dose has first been evaluated by a common thermoluminescence (TL) measurement, the same dosimeter is subjected to the PTTL dose reassessment with the intention of obtaining a new estimation by a completely independent measurement procedure. Recently, some experimental data indicated that PTTL could be used as a tool for dose discrimination in a mixed field using single TL detector. In this work TL (main dosimetry peak and high temperature peak) and PTTL characteristics for TLD-100 and TLD-700 detectors were investigated for Pu–Be neutrons. The results show that it is possible to determine separately gamma and neutron dose components with a single detector due to different sensitivities of TL and subsequent PTTL to neutrons in comparison to gamma rays. The main disadvantage, the low sensitivity of both TL and PTTL, to neutrons restricts the use only for the case of radiation accident.

SU-DD-A1-04: Monte Carlo Validation of Clinical Brachytherapy Dosimetry Under Partial Scatter Conditions for Neutron-Emitting Sources

Purpose: Monte Carlo (MC) models were generated in support of a clinical trial on the effectiveness of neutron-based brachytherapy for a patient treated with a plaque containing Cf-252 sources. Because the AAPM brachytherapy dosimetry formalism does not replicate partial scatter conditions of superficial brachytherapy, MC simulations were performed to evaluate treatment time and dose distributions generated using conventional methods. Method and Materials: Clinical calculations employed the AAPM dosimetry formalism with modified parameters for the neutron dose component. MC simulations utilized MCNP5 and track length estimator tallies. Computations applied a rectilinear mesh to tabulate neutron transport, including induced photons, and primary photon transport in a 14×14×5 cm3 volume with 9 mm3 voxels. Patient surface was simulated using a 20 cm radius hemisphere of water, with a corresponding hemisphere of air. For comparison to the AAPM formalism, the air was replaced with water. An RBE of 6 converted results to cGy-eq for the neutron component. Results were normalized to 0.1 mg Cf-252 source strength. Results: At the 6 mm prescription depth, calculated dose rates were 186±2 and 205±2 cGy-eq h−1 at plaque center and 24 mm offset, respectively. The central 4×4 cm2 area received 227±32 cGy-eq h−1. For comparison, full-scatter simulations yielded 205±2 cGy-eq h−1 at plaque center and 244±32 cGy-eq h−1 over a 4×4 cm2 area; although, computation time increased by a factor of 6.6. Dose ratios of full- (4π) to partial-scatter (2π) environments increased from 1.07 to 1.10 as depth increased from 0.4 to 5 cm. Approximately 90% of the dose-equivalent was due to neutrons, while neutron physical dose was 68% and 57% of the total at 0.6 and 5.0 cm depths, respectively. Conclusion: Dose can be overestimated upto 10% by assuming full-scatter conditions for Cf-252 plaque brachytherapy. MC simulations are recommended to validate treatment plans generated using conventional methods.

Set-up and calibration of a triple ionization chamber system for dosimetry in mixed neutron/photon fields

The aim of this study is to introduce a triple ionization chamber system to separate dose components of mixed neutron/photon fields. Fast and thermal neutron dose components have a different biological effectiveness than gamma dose components. If boron neutron capture is used to enhance the dose in certain areas of a patient, the precise knowledge of the thermal neutron flux is essential. A tissue equivalent and two magnesium ionization chambers have been prepared for use in a triple chamber system for this purpose. One of the magnesium chambers is coated with 10B on the inside to enhance its response to thermal neutrons. All three chambers have been calibrated at a cobalt source, medical linear accelerators and several neutron sources. The chambers have been studied in Monte Carlo simulations and the results are compared with measurements.

Method for determination of gamma and neutron dose components in mixed radiation fields using a high-pressure recombination chamber

A new method is proposed for the determination of dose components in mixed radiation fields (gamma + neutrons) using a recombination chamber. The method involves the determination of the ratio of ionisation currents measured at two different voltages applied to the chamber without the need of determining the saturation current, neither in the radiation field investigated nor during calibration. Therefore, the chamber can be filled with a gas under a pressure much higher than that used in presently available recombination chambers. This paper presents theoretically derived formulae supporting the method and the experimental results of dose component measurements using a high-pressure recombination chamber filled with methane. The method can be used for determining neutron and gamma dose components in the environment, especially in the vicinity of nuclear centres.

Radionuclide neutron sources in calibration laboratory--neutron and gamma doses and their changes in time

The calibration laboratory, having standard neutron fields of radionuclide sources, should perform regular measurements of fields' parameters in order to check their stability and to get knowledge of any changes. Usually, accompanying gamma radiation is not of serious concern, but some personal dosemeters, old neutron dose equivalent meters with scintillation detectors and the dose meters of mixed radiation require the determination of this component. In the Laboratory of Radiation Protection Measurements in the Institute of Atomic Energy, Poland, the fields of radionuclide neutron sources (252)Cf, (241)Am-Be and (239)Pu-Be were examined for nearly 20 y. A number of detectors and methods have been applied for the determination of neutron ambient dose equivalent rate and for the determination of neutron and gamma dose components. This paper presents the recent results of measurements of gamma and neutron dose and dose equivalent, compared with the results accumulated in nearly 20 y.

TL and PTTL of TLD-100 and TLD-700 after irradiation with 14.5 MeV neutrons

In radiation dosimetry, phototransfered thermoluminescence (PTTL) found its use for dose reassessment. After the dose has first been evaluated by a normal thermoluminescence (TL) measurement, the same dosimeter is subjected to the PTTL dose reassessment procedure with the intention of obtaining a new estimation by a completely independent measurement procedure. Recently, some experimental data indicated that PTTL could be used as a tool for dose discrimination in a mixed field using single TL detector. In this work, the short overview of the use of LiF-based TL detectors in mixed field dosimetry is presented. In the experimental part, TL (main dosimetry peak and high-temperature peak (HTP)) and PTTL characteristics for TLD-100 and TLD-700 detectors were investigated for 137Cs gamma rays and neutrons of 14.5 MeV. The results show that it is possible to determine separately gamma and neutron dose components with a single detector due to different sensitivities of TL and subsequent PTTL to neutrons in comparison to gamma rays.