Neutron Capture Gamma Rays Research Articles

Experimental validation of a spectroscopic gamma-ray detector based on a LaBr3 scintillator towards real-time dose monitoring in BNCT

We report the experimental validation of a detection module based on a LaBr3(Ce＋Sr) scintillator crystal, read by a matrix of Silicon Photomultipliers (SiPMs), for the development of a SPECT system for dose monitoring in Boron Neutron Capture Therapy (BNCT). The goal of the system is to detect the gamma rays at 478keV emitted by the excited 7Li produced in the 10B(n, α)7Li reaction, to have a real-time localization and quantification of the local dose released to a patient during BNCT treatment. The good energy resolution of the detector (<3% at 662keV) allows to resolve the photopeak at 478keV from the adjacent annihilation photons photopeak at 511keV, originated from pair production of high energy gamma rays, such as the hydrogen neutron capture gamma rays at 2.2MeV. The results shown in this article demonstrate the suitability of our detector in view of future application in BNCT real-time dose monitoring. Good linearity (RMSE < 2.5cps) between the number of events detected at 478keV and the boron concentration of the 10B-loaded samples has been achieved, down to a boron concentration of 62ppm with a neutron flux of approximately 1 × 105 n/cm2/s.

Analysis of the Impact of Cross-sectional Data Discrepancies on the Effectiveness of Radiation Shielding Design Using Monte Carlo Codes

The effect of cross-sectional data discrepancies on the effectiveness of radiation shielding design has been investigated in this paper. The MCNP code with cross-sections from ENDF/B-V and ENDF/B-VII has been used to determine the gamma ray dose equivalent, H*(10), behind a lead glass shield enclosing a slow neutron source. It is observed that the radiative capture gamma ray dose behind the shield is higher when ENDF/B-V cross-sections are used compared to that produced by ENDF/B-VII cross-sections. The discrepancy is due to absence of energetic primary gamma rays when ENDF/B-VII is used. The results show that shielding design using ENDF/B-VII cross-sections could underestimate the shield by a fair margin and compromise safety. It is therefore necessary to consider more than one release of ENDF/B when using lead glass for shielding slow neutron capture gamma rays. The discrepancies need to be addressed in the next releases.

CdZnTe quasi-hemispherical detector for gamma–neutron detection

ABSTRACT Quasi-hemispherical CdZnTe detector was manufactured successfully to fully understand the performance in the mixed gamma–neutron detection field. Together with the software of COMSOL, Geant4, and Matlab, the detector structure has been optimized. The CdZnTe detector performs good energy resolutions for 241Am, 57Co, and 137Cs radiation sources, especially for 137Cs (10.91 keV full width at half maximum [FWHM] at 662 keV). A linear relationship between the energy positions and spectrum channels indicates that the detector is effective for the precise energy detection from 59.5 to 662 keV. Finally, neutron and gamma events were detected simultaneously at room temperature using 241AmBe neutron source. The spectrum shows good energy resolution for neutron capture gamma ray (14.28 keV FWHM at 558 keV). Our work demonstrates that the quasi-hemispherical CdZnTe detector is promising for simultaneous detection of neutrons and gamma radiation.

Neutron dose in and out of 18 MV photon fields

In radiation therapy, neutron contamination is an undesirable side effect of using high energy photons to treat patients. Neutron contamination requires adjustments to the shielding requirements of the linear accelerator vault and contributes to the risk of secondary malignancies in patients by delivering dose outside of the primary treatment field. Using MCNPX, an established Monte Carlo code, manufacturer blueprints, and the most up to date ICRP neutron dose conversion factors, the neutron spectra, neutron/photon dose ratio, and the neutron capture gamma ray dose were calculated at different depths and off axis distances in a tissue equivalent phantom. Results demonstrated that the neutron spectra and dose are dependent on field size, depth in the phantom, and off-axis distance. Simulations showed that because of the low neutron absorption cross section of the linear accelerator head materials, the contribution to overall patient dose from neutrons can be up to 1000 times the photon dose out of the treatment field and is also dependent on field size and depth. Beyond 45cm off-axis, the dependence of the neutron dose on field size is minimal. Neutron capture gamma ray dose is also field size dependent and is at a maximum at a depth of about 7cm. It is important to remember that when treating with high energy photons, the dose from contamination neutrons must be considered as it is much greater than the photon dose.

NEUTRON CAPTURE GAMMA RAY FIELD WITH ENERGY TO 10 MeV FOR METROLOGICAL SUPPORT OF RADIATION PROTECTION DEVICES

Medical, and technological linear particle accelerators, and nuclear reactors are vastly widespread worldwide today. These facility generate fields of secondary gamma radiation with energy to 10 MeV. Therefore, we have a need to calibrate spectrometric and dosimetric ionization measurement instruments for the energies to 10 MeV. The aim of this work is to determine possibility to use thermal neutron collimator of АТ140 Neutron Calibration Facility with 238Pu-Be fast neutron source (IBN-8-6) for this. Below 3 MeV we use a set of point gamma standard spectrometry sources OSGI. We can acquire gamma rays with energies above 3 MeV using radioactive thermal neutron capture on target, i.e. (n, γ)-nuclear reaction. We can use neutron capture gamma-ray from titanium target (to 7 MeV) or nickel target (to 10 MeV) situated in thermal neutron field for calibration. We can use thermal neutron collimator of АТ140 Neutron Calibration Facility with 238Pu-Be fast neutron source (IBN-8-6) for slowing down neutrons from radionuclide fast neutron sources to thermal energies in polyethylene. Thermal neutron collimator forms a beam from radionuclide source with a significant amount of neutrons with thermal energies. We placed Ti and Ni targets in collimator’s canal. We got experimental spectral data on detection unit BDKG-19M NaI(Tl) 63 × 160 mm with nonlinear channel-energy conversion characteristic in range to 10 MeV. For additional filtration we proposed to use polyethylene neutron reflector and lead discs. We experimentally determined that placement of lead discs in collimator in front of the target allows to filter all spectrum while insignificantly weakening target’s emission. Using theoretical and experimental data we proved the ability to calibrate gamma-ray spectrometers in the range to 10 MeV.

Measurement of neutron capture cross sections and neutron capture gamma-ray spectra of 93Nb, 103Rh and 115In in keV-energy region

The neutron capture cross sections of 93Nb, 103Rh and 115In were measured by the time-of-flight method in the energy region from 15 keV to 100 keV. A pulse-height weighting technique was employed to derive the neutron capture cross sections. The neutron capture γ-ray spectra were derived by unfolding the pulse-height spectra with the detector response functions.

Study on the keV neutron capture reaction in 56Fe and 57Fe

The neutron capture cross-sections and the radiative capture gamma-ray spectra from the broad resonances of 56Fe and 57Fe in the neutron energy range from 10 to 90keV and 550keV have been measured with an anti-Compton NaI(Tl) detector. Pulsed keV neutrons were produced from the 7Li\( (p,n)\) 7Be reaction by bombarding the lithium target with the 1.5ns bunched proton beam from the 3MV Pelletron accelerator. The incident neutron spectrum on a capture sample was measured by means of a time-of-flight (TOF) method with a 6Li -glass detector. The number of weighted capture counts of the iron or gold sample was obtained by applying a pulse height weighting technique to the corresponding capture gamma-ray pulse height spectrum. The neutron capture gamma-ray spectra were obtained by unfolding the observed capture gamma-ray pulse height spectra. To achieve further understanding on the mechanism of neutron radiative capture reaction and study on physics models, theoretical calculations of the \( \gamma\) -ray spectra for 56Fe and 57Fe with the POD program have been performed by applying the Hauser-Feshbach statistical model. The dominant ingredients to perform the statistical calculation were the Optical Model Potential (OMP), the level densities described by the Mengoni-Nakajima approach, and the \( \gamma\) -ray transmission coefficients described by \( \gamma\) -ray strength functions. The comparison of the theoretical calculations, performed only for the 550keV point, show a good agreement with the present experimental results.

Quantitative analysis of neutron-capture gamma-ray energy spectra using direct demodulation

Because of the effects of background and measurement environment and multiplet effects of different elements, high-precision analysis of mixed capture [Formula: see text]-ray energy spectra of complicated formations remains challenging for geochemical elemental logging. The direct demodulation (DD) method makes full use of the measured data information, enabling physical constraints to be rationally applied to the spectral analysis process, and can yield high-precision elemental content from poor-statistics, low signal-to-noise ratio, and disturbed data. We construct mixed formations of different sandstones and limestones, mixed formations of sandstone and anhydrite, and more complicated mixed formations of multiple lithologies and employ Monte Carlo numerical simulations to obtain the neutron-capture [Formula: see text]-ray energy spectra of these mixed formations. We then employ the DD method and the weighted-least-squares (WLS) method to analyze quantitatively such mixed spectra, respectively, and compare the results with the actual contents of formation elements. The results indicate that the DD method offers higher precision spectral analysis compared with the results of the WLS method. The results for the capture [Formula: see text]-ray energy spectra of the formation for two actual wells also indicate that the DD method can be useful for spectral analysis in actual application.

Measurements of keV-neutron capture cross sections and capture gamma-ray spectra of 105Pd

The neutron capture cross sections and capture gamma-ray spectra of 105Pd were measured in the region from 15 to 100 keV and at 585 keV. A neutron time-of-flight method was utilised with an anti-Compton NaI(Tl) spectrometer and a 1.5-ns pulsed neutron source by the 7Li(p,n)7Be reaction. The capture yields were obtained by applying a pulse-height weighting technique to the observed net capture gamma-ray pulse-height spectra. The capture cross sections of 105Pd were derived with errors less than 5%, using the standard capture cross sections of 197Au. The evaluated capture cross sections of JENDL-4.0 and ENDF/B-VII.1 were compared with the present results. The evaluations of JENDL-4.0 and ENDF/B-VII.1 were larger than the present results by 3%–15% in the region from 15 to 100 keV and at 585 keV. The capture gamma-ray spectra of 105Pd were also derived by unfolding the observed net capture gamma-ray pulse-height spectra. The multiplicities of capture gamma rays of 105Pd were obtained from the capture gamma-ray spectra.

Determination of wax deposition and corrosion in pipelines by neutron back diffusion collimation and neutron capture gamma rays

Wax deposition in pipelines can be very costly for plant operation in oil industry. New techniques are needed for allocation and thickness determination of wax deposits. The timely removal of wax can make large saving in operational cost. Neutron back diffusion and neutron capture gamma rays were used in this study to measure paraffin, asphalt and polyethylene deposition thicknesses inside pipes and to enable simultaneous determination of scale and pipe corrosion. It was possible to determine a thickness change of less than one mm in 2min. It was also possible to detect localized scale from a small region of the pipe of approximately 2cm in diameter. Although experiments were performed in lab, the system can be made portable for field applications.

A new analytical formula for neutron capture gamma dose calculations in double-bend mazes in radiation therapy

BackgroundPhotoneutrons are produced in radiation therapy with high energy photons. Also, capture gamma rays are the byproduct of neutrons interactions with wall material of radiotherapy rooms. AimIn the current study an analytical formula was proposed for capture gamma dose calculations in double bend mazes in radiation therapy rooms. Materials and methodsA total of 40 different layouts with double-bend mazes and a 18MeV photon beam of Varian 2100 Clinac were simulated using MCNPX Monte Carlo (MC) code. Neutron capture gamma ray dose equivalent was calculated by the MC method along the maze and at the maze entrance door of all the simulated rooms. Then, all MC resulted data were fitted to an empirical formula for capture gamma dose calculations. Wu–McGinley analytical formula for capture gamma dose equivalent at the maze entrance door in single-bend mazes was also used for comparison purposes. ResultsFor capture gamma dose equivalents at the maze entrance door, the difference of 2–11% was seen between MC and the derived equation, while the difference of 36–87% was found between MC and the Wu–McGinley methods. ConclusionOur results showed that the derived formula results were consistent with the MC results for all of 40 different geometries. However, as a new formula, further evaluations are required to validate its use in practical situations. Finally, its application is recommend for capture gamma dose calculations in double-bend mazes to improve shielding calculations.

Investigation the Effects of Neutron Source Number and Arrangement in Landmines Detection by Thermal Neutron Capture Gamma Ray Analysis

Thermal neutron capture can be used as a successful technique for detection of non-metallic landmines via the detection of their constituent like nitrogen. Recently, it has been shown that the detection of 10.829 MeV photons from the 14N(n,γ)15N reaction can be used for finding the landmines. In this method a high-energy neutron source like 241Am-Be inside water as a moderator is used to have thermal neutron. In this paper we have investigated the effects of the number of neutron sources and their orientation on the gamma ray spectrum by using MCNP4C code. The best case for number of sources and their positions and orientations have been achieved corresponding to maximum flux of 10.829 MeV photons.

Erratum to: Determination of k 0-values for some elements in neutron capture gamma-ray for prompt gamma activation analysis

Exploration and modification of data for k 0-PGAA (prompt γ-ray activation analysis) values used as standards for k 0 standardization are needed. An integrated system has been installed and calibrated in Hot Laboratories Center for PGAA, using 252Cf isotopic neutron source with neutron flux of 6.16E8 n/cm2 s. The prompt k 0-factors of about 12 elements were determined versus the 1951.1 keV γ-ray of the 35Cl, as well as analyzing crude oil and oil product samples. The concentrations of the elements in crude oil H, B, Mg, Al, Na, Si, P, S, Cl, V and Cd were 1.18E5, 0.084, 5.48E2, 8.45E2 4.88E2, 1.62E2, 3.72E2, 8.2, 144.3, 393 and 209.2 ppm while in oil product samples were 1.32E5, 5.87, 4.56E2, 4.22E2, 7.16E3, 699, 157, 8.74, 492.3, 61 and 198.2 ppm. Validation of the k 0-PGAA was applied in analyzing standard IAEA reference material (soil-7) which give good agreement with the literature data. The 252Cf neutron beam in radiation position was characterized by the cadmium-ratio method, and found that the cadmium ratio (f), was 160.

Evaluation of Gd, Sm and Eu in Egyptian monazite samples using the prompt neutron capture gamma-ray technique

Three Egyptian monazite samples from the beach black sand at Abou Khashba near Rosetta, submitted by the laboratories of the nuclear materials authority of Egypt were analyzed for elemental contents, using Prompt gamma ray neutron activation analysis (PGNAA) as an effective analysis technique especially for rare earth elements. An integrated system for PGNAA, located at hot Lab. Center, has proven useful for the analytical purpose using 252Cf neutron source with neutron flux of 6.16E7 n/cm2 s, that system has been designed and calibrated for PGNAA. Sensitivity curves using different standard concentration values of artificial liquid standards samples at different concentrations ranging from 1,000 to 5,000 mg/L in case of Sm and Gd, while from 100 to 1,000 mg/L for Eu, have been constructed. The average concentration values for Gd, Sm and Eu in the samples under investigation was at 1.46E4 ± 0.21, 6.683E4 ± 0.14 and 440 ± 0.18 ppm, respectively. A comparative study of the obtained results and the results of ICP-Ms was given.

Measurement of Neutron Capture Gamma Rays from the Resonances of 91Zr and 96Zr at the J-PARC/MLF/ANNRI

Measurement of Neutron Capture Gamma Rays from the Resonances of 91Zr and 96Zr at the J-PARC/MLF/ANNRI

A CdTe detector with a Gd converter for thermal neutron detection

Abstract A CdTe detector with a Gd converter has been developed and investigated as a neutron detector for neutron imaging. The fabricated Gd/CdTe detector with the 25 μm thick Gd was designed on the basis of simulation results of thermal neutron detection efficiency and spatial resolution. The energy resolution of the Gd/CdTe detector is less than 4 keV, which is enough to discriminate neutron capture gamma rays from background gamma emission. The Gd/CdTe detector shows the detection of neutron capture gamma ray emission in the 155 Gd(n, γ) 156 Gd, 157 Gd(n, γ) 158 Gd and 113 Cd(n, γ) 114 Cd reactions and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application.

Neutron Detection With Gamma-Ray Spectrometers for Border Security Applications

Development of technologies for neutron detection that do not require 3He is important because the supply of 3He is very limited, and the cost of the gas is becoming prohibitive for many applications. This study evaluates the ability to detect neutron sources with gamma-ray spectrometers that are already present in many radiation measurement systems. Detection is based on count rates for gamma rays in the 3 to 8 MeV range, which are produced by the emission of fission gamma rays and neutron capture reactions in vehicles and their cargo. For materials in the normal stream of commerce, gamma rays above 3 MeV are produced only by sources that also emit neutrons. Therefore, unless the gamma-ray count rate is high enough to produce excessive random pileup, the detection of high-energy gamma rays provides an unambiguous indication of the presence of a neutron source. As part of this investigation, several shields that are suitable for use in radiation portals were constructed and characterized for their abilities to produce additional high-energy, neutron-capture gamma rays. A shield (composed of alternating layers of polyethylene and steel) enhances the ability to detect neutrons without producing detrimental effects for gamma-ray measurements. Calculations show that when shielded by neutron-detection-enhancing materials, NaI detectors can be as sensitive to the presence of a concealed neutron source as moderated 3He detectors.

Se-78,80,82の550 keV中性子捕獲断面積と捕獲γ線スペクトルの測定

Se-78,80,82の550 keV中性子捕獲断面積と捕獲γ線スペクトルの測定

Absolute efficiency calibration of high-purity germanium detector in the range 120–8500 keV

This article reports the efficiency response curve of the high-purity germanium detector over the wide energy range, covering from 120 to 8500 keV. The efficiencies were measured for different counting geometries by using point radionuclide standards (mono-energetic as well as multi-gamma emitters) supplied by IAEA and the capture gamma-ray facility installed at PINSTECH nuclear reactor PARR-1. The measured efficiencies were required to fit with a suitable fitting function for interpolation within the energy range of interest. Several fitting functions were proposed in the literature covering different energy ranges. The functions giving the best fit to experimental data are presented. The work has successfully extended the response curve beyond 1500–8500 keV, which is the region where the standard calibration radionuclides are not available. The thermal neutron capture gamma-ray facility provided the collimated neutron beam, extracted from the core of the reactor and made to react with ammonium chloride target to produce the capture gamma rays for determining the efficiencies in the extended region. It was found that the capture gamma-ray provides a satisfactory solution to extend the absolute efficiency calibration in the MeV range. It was also found that the fitting function that is linear in its parameter was highly satisfactory up to 1500 keV but proved insufficient upto 8500 keV. The exponential function giving the good fit over the range has been presented. Good agreement has been found between the experimentally measured absolute efficiencies and the predicted result.

Errors, PGAA, and metrology

Analytical chemists are scientific scavengers, using any available physical phenomenon to determine the composition of materials. Accuracy, unlike mere reproducibility, depends on understanding the measurement process, most usefully by stressing it until it bends. The use of neutron capture gamma rays for analysis involves some issues not encountered in conventional neutron activation analysis which, when understood, extend the utility and reinforce the value of nuclear methods of elemental analysis. These methods are being increasingly recognized as tools for SI-traceable metrology.