Am-Be Neutron Source Research Articles

Predicting scale thickness in three-phase flow using neutron activation analysis and deep learning

Calcium carbonate scales occur on the inner walls of the pipes used in oil and gas production, reducing the internal diameter and obstructing fluid flow. This study proposes a method for predicting of scale thickness using a neutron source and gamma radiation analysis with a deep neural network (DNN). The detection system includes a 241Am-Be neutron source and 72 detectors that count particles and record photons. A measurement setup was designed using the MCNP6 code to optimize the detection of scale thickness, utilizing a flat neutron source and strategically positioned detectors. The gamma-ray spectra generated from this setup were used as input data for the DNN to predict scale thickness within a three-phase water–gas-oil annular system. The results indicated that the DNN accurately predicted scale thickness with a mean relative error of 3.01 %, achieving a relative error below 5 % for 92.68 % of the dataset, demonstrating the potential for reliable real-time monitoring in industrial settings.

Characterization of neutron spectrum of polyethylene-moderated AmBe neutron source using a passive single-cylindrical neutron spectrometer

The changes in the neutron spectrum produced by the AmBe neutron source moderated in the polyethylene sphere need to be characterized. A passive single-moderator neutron spectrometer with indium foil activation neutron detector has been developed to characterize the neutron spectrum of an AmBe source inside a polyethylene sphere. The detector response function was calculated using the MCNPX 2.7 program with IRDFF-II nuclear data library. Measurements were conducted by placing the Single-Cylindrical Neutron Spectrometer (SCNS) on the outer surface of the sphere for 5 h. The 116mIn activity due to neutron activation in each indium foil was measured using a gamma spectrometer with an HPGe detector. By identifying the count value at the peak energy of 1294 KeV and considering an HPGe detector efficiency of 3.2% at the foil position 1 mm above the detector surface, the activity of 116mIn was obtained. The activity value of 116mIn from each indium foil was compared with the MCNPX simulation results. The neutron spectrum was unfolded using the UMG 3.3 program with activity data input of 116mIn for each foil and detector response. A neutron spectrum was obtained with a total neutron flux of 634 ± 60 n/cm2·s, consisting of 25% thermal neutrons, 16% epithermal neutrons, and 61% fast neutrons. When compared with the simulation results, the total neutron flux in the spectrum produced by SCNS-In showed only a small difference of 1%. Based on these neutron spectrum measurements, it was determined that placing the AmBe neutron source inside a 15″ diameter PE-sphere will reduce the fast neutron flux by 78%.

Design and Development of a Portable Neutron Detection System Based on a 6Li-Doped ZnS(Ag) Scintillator

In this study, the design and development of a portable neutron detection system based on a 6LiF + ZnS(Ag) scintillator was carried out. The detector system was first modeled using the Monte Carlo simulation code MCNP. As a result of the detector simulation, the materials and photomultiplier tubes (PMTs) were supplied for the system, and the material properties and thicknesses were optimized for the best scintillator, with the EJ-420 (Eljen Technology), consisting of 95% 6Li dispersed in ZnS(Ag), chosen because it has a distinctive feature with regard to high thermal neutron detection efficiency. Since this material has been shown to offer a significant advantage in favor of neutron detection in mixed radiation fields with gammas and neutrons, the pulse shape discrimination method was employed. To achieve this, a proper electronic circuit was developed to discriminate the pulses from neutrons and gammas. In the experimental step, the EJ-420 scintillator with a 50-mm diameter was optically coupled with a special fast PMT (Hamamatsu H1949-51 model). In conclusion, this study, which underlines the performed simulations for a neutron detector configuration, gives the obtained experimental results showing discrimination capability in neutron/gamma detection using a 241Am-Be neutron source. The results show that the EJ-420 is a good scintillator due to its highly enriched 6Li transmator, which results in more effective neutron measurements when a portable neutron detector design is chosen.

Measurements of the Neutron Field Parameters with Monitors based on the Gas-Filled Proportional Counters

Neutron monitors with the gas filled proportional counters inside and first data were presented at RuPAC-2018 and RuPAC-2021 conferences. It is planned to use such devices to monitor the stability of irradiation conditions during radiation therapy sessions. We present here the calibration procedure for these monitors at AmBe neutron source, and the new data. These monitors have been used to measure the neutron fluence behind the shielding of the 450 MeV/nucleon carbon beam based experimental facility “Shared use Center – radiobiological stand with the carbon beam of the U-70 accelerator at IHEP”. The measurements were supported by the set of extensive simulations using the CERN FLUKA code. Good agreement between the FLUKA simulations and measurements was shown. Even single monitor with lead insertion calibrated at AmBe neutron source allows to have a good result in the real neutron field. The possibility to estimate the fluence of the high energy neutrons with the energy above 10 MeV using data from the pair of monitors was demonstrated. To improve the quality of the measurements one needs to take into account difference between calibration and measurements conditions.

Diamond based neutron detector in high temperature environments of 200 °C

The stability of the neutron response function and the detection efficiency of a sCVD diamond detector in the temperature range up to 200 °C are demonstrated in this paper. A CIVIDEC neutron detector was simultaneously heated with an AmBe neutron source. This enabled the measurement of the spectrum of the deposited energy in the diamond sensor, i.e. the neutron response function of the detector and the detection rate, which in combination represents the neutron flux as a function of temperature. The measured temperature stability of the neutron response function of the detector and detection rate demonstrates the suitability of sCVD diamond detectors for harsh environments, such as encountered in geodetic applications and nuclear fusion research.

Effect of fibers and boron carbide on the radiation shielding properties of limestone and magnetite aggregate concrete

Research on new composite materials in the field of radiation shielding materials requires adequate means and meticulousness in their manufacture and characterization. In this study, concretes that use elements that theoretically improve their shielding properties are developed, such as fibers and boron-rich additions. Both magnetite and limestone concrete have been manufactured, incorporating steel fibers to improve the linear attenuation coefficient and polyvinyl alcohol to increase neutron scattering cross section. The incorporation of boron carbide is intended to increase the neutron absorption cross section. 5 mixes have been manufactured that have been tested with several gamma radiation sources (570–1280 keV) and with an Am–Be neutron source (111 GBq). The experimental results have been used to validate simulations carried out with MAVRIC® and have led to the modelling of a simplified spent nuclear fuel cask. Boron carbide produces a decrease of more than 50% on the neutron dose rate in contact with the cask. Incorporating PVA fibers supposes gain in neutron attenuation capacity of up to 13%, while fibers have a negligible effect on the photon dose rate.

Calibration and validation measurements of the Extended-range Bonner sphere spectrometer

An Extended-range Bonner sphere spectrometer (EBSS) has been developed to investigate the neutron spectra of the China initiative Accelerator Driven System. This paper presents the design, calibration, and validation measurements of the EBSS system using a standard 241Am-Be neutron source and the cosmic ray neutrons. The EBSS system was simulated using the PHITS code, and the geometric structures were designed using the response functions obtained by simulations. The EBSS system comprises of seven polyethylene-only spheres and seven extended-range spheres encased in lead, copper, or tungsten shells along with a bare 3He proportional counter. To verify the accuracy of the simulated response functions, the EBSS system was calibrated with monoenergetic neutron beams of 565 keV at the China Institute of Atomic Energy in Beijing. The EBSS system was used to measure neutron count rates of the 241Am-Be neutron source and cosmic ray neutrons. Subsequently, the measured neutron count rates were converted into neutron spectra by the unfolding software. Experimental findings demonstrate the precise capability of the EBSS system to measure neutron spectra across a wide range of neutron radiation fields.

Long-term irradiation of an ATLAS NSW SM2 Micromegas quadruplet using an AmBe neutron source

The NSW Micromegas chambers in the ATLAS forward muon spectrometer are subject to background rates of 15-20kHz/cm2 under HL-LHC conditions. The innermost detector area closest to the LHC beam pipe will accumulate a charge of 0.068Ccm−2year−1 under these rates. Due to the late change of the detector gas from non-aging Ar:CO2 93:7vol% to the more HV stable ternary mixture Ar:CO2:iC4H10 93:5:2vol% and the known vulnerability of wire chambers to Hydrocarbon-containing gas mixtures a three-year-long aging study has been performed. An SM2 series module of the NSW Micromegas quadruplets was irradiated at LMU in Garching/Munich using a 10GBq AmBe neutron source emitting 6×105MeV n/s as well as 3.5×105 4.4MeV gammas/s and 3.6×109 60keV gammas/s. The SM2 chamber was irradiated in a region of several 10cm2 in size with a dose rate well exceeding the HL-LHC equivalent local charge densities for three years. In between the irradiation periods the performance of the SM2 chamber regarding spatial resolution and efficiency on cosmic muon tracking was tested several times. We report on the irradiation and the performance studies of the SM2 Micromegas quadruplet and conclude that no sign of loss in performance has been observed in contradiction to an earlier experience using drift tube wire chambers.

Design and characterization of a new high-count wide-range fission chamber based on GEM

As one of the main detectors for monitoring neutron flux rate in a nuclear reactor, the fission chamber (FC) suffers from issues such as low sensitivity, limited counting rate dynamic range, and system mode switching is cumbersome. This study utilizes the advantages of gas electron multipliers (GEM), which are easy to fabricate in large areas and have high counting rates, to design a novel fission chamber. By conducting Monte Carlo simulations on parameters such as the thickness of the 235U coating, the drift distance, and the operating electric field strength, a wide-range fission chamber design that combines high sensitivity and a counting rate range based on GEM has been obtained. The simulation results demonstrated that using a GEM detector to build a fission chamber can significantly improve sensitivity and extend the counting rate range. Subsequently we built a proof-of-concept GEM-based fission chamber and tested it with a 55Fe low-energy X-ray source and an Am-Be neutron source. The results showed that the proof-of-concept detector had a good signal-to-noise ratio and energy linearity, as well as clear discrimination between alpha background and neutron pulse amplitudes.

Development of a silicon-based thermal neutron detection system

Neutron detection systems are of increasing importance in applications from basic science to medical applications and reactors. 3He proportional counters remain the most popular choice for monitoring thermal neutrons with a detection efficiency of around 60%, however, due to 3He global shortages, a new generation of detection technologies will be required to meet the rising demand. As a result, extensive research is being conducted to investigate alternative methods of neutron detection. This work presents such a system and demonstrates its calibration and evaluation using an AmBe neutron source. The detection system involves silicon sensors coated by converter layers to make the detectors sensitive to thermal neutrons via neutron capture and measurement of the resulting secondary charged particles. The detection system is presented in two configurations, a single and a multi-layer configuration, where the latter is used to increase the total detection efficiency. In addition, the system is capable of determining coincident signals from a single neutron capture, a feature which allows background suppression and an increase in the purity of the neutron signal which is particularly useful in mixed radiation environments.

Effect of ZrO2 on physical and radiation shielding properties of tellurium −lead- calcium −borate (TPCB)glass system

This research investigates the impact of adding Zirconia (ZrO2) to tellurium-lead-borate glasses, focusing on how it affects their physical, optical, and radiation shielding properties, particularly against gamma-rays and neutrons. Key parameters for gamma-ray shielding (mass attenuation coefficient (µm), effective atomic numbers (Zeff), and effective electron densities (Neff)) were measured across an energy range of 0.080 to 2.614 MeV. Notably, excellent agreement (within 0.5 %) was observed between the measured and calculated µm values. For slow neutron attenuation measurements, we employed a 241Am-Be neutron source in conjunction with a BF3 detector. Fast neutron attenuation was assessed using the NXCOM software to determine the effective macroscopic removal cross-section (ΣR). The addition of Zirconia notably enhanced the glass density (∼6.8 %) and its ability to shield against gamma-rays (∼10 %) and slow neutrons (∼71 %). However, this improvement comes with a slight trade-off: a decrease of approximately 1.5 % in fast neutron attenuation was observed. This trade-off between shielding effectiveness for different radiation types warrants further investigation to optimize glass compositions, considering the desired balance between shielding performance across various radiation spectrums.

Design and validation of a single cylindrical neutron spectrometer using a gold activation foil

A single cylindrical neutron spectrometer (SCNS) with a gold activation foil was designed for mono-directional high-flux neutron spectrum measurement applications. A total of seven gold activation foils with a diameter of 11 mm and a thickness of 0.2 mm were placed in the moderator as detectors with a depth variation of 15 mm. The moderator was cylindrical with a diameter of 8 cm and a length of 16 cm and made of polyethylene. Outside the moderator is a neutron shield made of 3 mm boron-rubber and 0.5-mm cadmium and covered with an aluminum casing of 5 mm thickness. The response value of each detector was determined by simulation using the Monte Carlo MCNPX 2.7 program. The simulated response detectors was then partially validated by comparing the simulation and the measurement result with an AmBe neutron source, where the difference is around 3%. The obtained results showed that SCNS with a gold activation foil could be used to measure the neutron spectrum in a high-flux neutron field.

Development of a High-Efficiency Device for Thermal Neutron Detection Using a Sandwich of Two High-Purity 10B Enriched Layers.

The shortage of 3He, a crucial element widely used as a neutron converter in neutron detection applications, has sparked significant research efforts aimed at finding alternative materials, developing appropriate deposition methods, and exploring new detector architectures. This issue has required the exploration of novel approaches to address the challenges faced in neutron detection. Among the available conversion materials, 10B has emerged as one of the most promising choices due to its high neutron-capture cross-section and relatively high Q value. In our previous papers, we delved into the possibility of depositing neutron conversion layers based on 10B using Pulsed Laser Deposition (PLD). We investigated and evaluated the performance of these layers based on various factors, including deposition conditions, substrate properties, and film thickness. Moreover, we successfully developed and tested a device that employed a single conversion layer coupled with a silicon particle detector. In this current study, we present the development of a new device that showcases improved performance in terms of efficiency, sensitivity, and discrimination against γ background signals. The background signals can arise from the environment or be associated with the neutron field. To achieve these advancements, we considered a new detection geometry that incorporates the simultaneous use of two 10B conversion layers, each with a thickness of 1.5 μm, along with two solid-state silicon detectors. The primary objective of this design was to enhance the overall detection efficiency when compared to the single-layer geometry. By employing this novel setup, our results demonstrate a significant enhancement in the device's performance when exposed to a neutron flux from an Am-Be neutron source, emitting a flux of approximately 2.2 × 106 neutrons per second. Furthermore, we established a noteworthy agreement between the experimental data obtained and the simulation results.

Combined anticancer effects of neutron radiation and genistein on prostate cancer cells

Prostate cancer is the most frequent cancer in men and the second largest cause of cancer-related mortality after lung cancer. Today, studies on neutron radiation therapy show that this method alone is not able to prevent the progression of the disease in most patients with prostate cancer, so it is better to combine neutron radiation with an auxiliary antitumor such as genistein. We found that genistein has anticancer properties and few antioxidant capabilities and could cause the death of prostate cancer cells, mostly through triggering apoptosis. Here the purpose of this study was to evaluate the effect of very low dose rate neutron irradiation emitted from 241Am-Be neutron source as well as the effect of the combination of neutron irradiation and adding the optimal concentration of genistein (480 μM) on the attenuation of prostate cancer PC3 cells in 3D culture environment. MTT assay, Neutral red assay, NO assay, Catalase activity, Reduced glutathione (GSH), Cytochrome c releasing, and Comet assay were used to assess the effect and type of mortality caused by neutron radiation and also neutron and genistein combination in prostate cancer cells. The results of cytotoxicity test reveal that very low dose rate neutron radiation (0.6 mGy/h) with or without genistein at 5, 10, 15, and 20h reduced the survival of prostate cancer cells. Moreover, the rate of necrosis due to the neutron effect at different irradiation times enhanced with the increasing time. Despite significant anticancer effect of Genistein, it cannot be used simultaneously with neutron radiation for prostate cancer treatment.

Assessment of uranium-235 distribution in soil samples from Anbar province, Western Iraq

Uranium is a radioactive element that can pose a risk to human health if ingested or inhaled. Exposure to high levels of uranium can increase the risk of cancer. Measuring uranium concentration in soil can help identify areas where there may be a risk of exposure to high levels of uranium and take appropriate measures to protect public health. In this study, uranium concentration in soil samples collected from Anbar province in Iraq was measured using CR-39 detectors irradiated with an Am-Be neutron source. The results showed that the highest concentration of uranium-specific activity (12.029 ± 0.0016 Bq kg−1) was found in Al Fallujah district, while the lowest concentration (0.531 ± 0.0007 Bq kg−1) was found in Al Sagra district. Interestingly, the data suggest that uranium concentrations increase as we move southeast towards Baghdad, where the city’s climate is reportedly polluted. Despite this, the average uranium-specific activity level in Anbar (7.668 Bq kg−1) is relatively low, and it falls below the global average of 35 Bq kg−1 recommended by the United Nations Scientific Committee on the Effects of Atomic Radiation. This study concludes that there is no evidence of uranium pollution in Anbar. Although uranium concentrations are considered low, low levels of uranium concentration in soil can have important implications for environmental risk, resource exploration, agricultural productivity, and geological research.

Measurement of flux distribution of an AmBe neutron source and estimation of two group integral capture cross-sections

Neutron capture cross sections of elements are of utmost importance in reactor physics, dosimetry applications and stellar nucleosynthesis studies. Manipal Centre for Natural Sciences (MCNS) possesses a 16Ci AmBe neutron source, which is housed inside a concrete structure. The neutrons undergo multiple scattering events in the concrete structure and surrounding materials; therefore, the neutron flux profile is unique to this source and structural material, which differs from the standard ISO spectrum. Using this AmBe source, experiments are being conducted to understand the neutron energy distribution in two groups to measure capture cross sections of various isotopes. In the present work, gold has been used as a flux monitor, and the indium cross-section has been measured in two groups. These foils (Gold and Indium) are irradiated with and without cadmium cover to understand the neutron spectra in two spectral regions as cadmium cutoffs neutrons of energy below 0.5 eV. Irradiated samples were counted in High-Pure Germanium (HPGe) detector, and the γ-spectra obtained were used to get the epi‑cadmium (0.5 eV and above) to total ratio of ⟨σ(E)ϕ(E)⟩. The neutron flux is estimated in two groups corresponding to thermal (thermal energy neutron extended up to 0.5 eV) and epi‑cadmium regions, and also the flux-weighted capture cross sections of 115In are determined.

Calibration strategy of the PROSPECT-II detector with external and intrinsic sources

This paper presents an energy calibration scheme for an upgraded reactor antineutrino detector for the Precision Reactor Oscillation and Spectrum Experiment (PROSPECT). The PROSPECT collaboration is preparing an upgraded detector, PROSPECT-II (P-II), to advance capabilities for the investigation of fundamental neutrino physics, fission processes and associated reactor neutrino flux, and nuclear security applications. P-II will expand the statistical power of the original PROSPECT (P-I) dataset by at least an order of magnitude. The new design builds uponprevious P-I design and focuses on improving the detector robustness and long-term stability to enable multi-year operation at one or more sites. The new design optimizes the fiducial volume by elimination of dead space previously occupied by internal calibration channels, which in turn necessitates the external deployment. In this paper, we describe a calibration strategy for P-II. The expected performance of externally deployed calibration sources is evaluated using P-I data and a well-benchmarked simulation package by varying detector segmentation configurations in the analysis. The proposed external calibration scheme delivers a compatible energy scale model and achieves comparable performance with the inclusion of an additional AmBe neutron source, in comparison to the previous internal arrangement. Most importantly, the estimated uncertainty contribution from the external energy scale calibration model meets the precision requirements of the P-II experiment.

Simultaneous detection of fast and thermal neutrons with a stilbene-6Li glass composite scintillator

For simultaneous detection of fast and thermal neutrons, a composite scintillator consisting of stilbene crystal and 6Li glass was designed. Energy calibration was finished with a 152Eu γ source through Compton coincidence measurement. A 241Am-Be neutron source with polyethylene moderator was used to test it. The detection efficiency was examined through Monte Carlo simulation. The measured results show that the composite scintillator has excellent triple pulse shape discrimination performance. With the threshold of 100 keVee, figure of merits for γ-fast neutron and fast neutron-thermal neutron are 1.571 ± 0.009 and 2.376 ± 0.010 respectively. Two thermal neutron peaks emerge at 0.245 MeVee and 0.291 MeVee in the energy spectra. 79.24% of the incident thermal neutrons can be detected by 6Li glass. 39.56% of 1 MeV neutrons can be detected by stilbene.

A large area, high counting rate micromegas-based neutron detector for BNCT

Beam monitoring and evaluation are very important to boron neutron capture therapy (BNCT), and a variety of detectors have been developed for these applications. However, most of the detectors used in BNCT only have a small detection area, leading to the inconvenience of the full-scale 2-D measurement of the beam. Based on micromegas technology, we designed a neutron detector with large detection area and high counting rate. This detector has a detection area of 288 mm × 288 mm and can measure thermal, epithermal, and fast neutrons with different detector settings. The BNCT experiments demonstrated that this detector has a very good 2-D imaging performance for the thermal, epithermal, fast neutron and gamma components, a highest counting rate of 94 kHz/channel, and a good linearity response to the beam power. Additionally, the flux fraction of each component can be calculated based on the measurement results. The Am-Be neutron source experiment indicates that this detector has a spatial resolution of approximately 1.4 mm, meeting the requirements of applications in BNCT. It is evident that this micromegas-based neutron detector with a large area and high counting rate capability has great development prospects in BNCT beam monitoring and evaluation applications.

THE IMPACT OF NEW ICRU95 QUANTITIES AND SPECTRUM DATA ON THE DOSIMETRIC QUANTITIES OF THE NEUTRON REFERENCE FIELDS AT PTB.

In 2020, the ICRU released a new report which includes the re-definition of the operational quantities used in radiation protection and new conversion coefficients from physical quantities to operational radiation protection quantities. An assessment of the ambient and personal dose conversion coefficients for the reference neutron fields of radionuclide sources at PTB is necessary based on these new definitions. In this work, a numerical estimation of the conversion coefficients of moderated and unmoderated 252Cf and 241Am-Be neutron sources based on ICRU57 and ICRU95 reports and using spectrum data available in the ISO 8529-1 standard and at PTB are discussed. Two numerical approaches are used for this estimation to ensure the reliability of the calculated values: a direct calculation using MCNP6, and cubic interpolation of conversion coefficients datasets written in Python. The results show large differences between the spectrum-averaged operational quantities for the current and new conversion coefficients of up to 23%. The choice of spectrum data affects conversion coefficient values by 6-8%.