Thermal Neutron Counting Research Articles

Research on burnup depth measurement of spent fuel board based on thermal neutron imaging technology

The detection equipment for measuring the burnup depth in spent fuel rods is a vital component of the spent fuel management system. which can determine the burnup depth of spent fuel board and plays a crucial role in safeguarding the safety, economic efficiency, and structural integrity of the fuel assembly. This study introduces an innovative technical approach for assessing the burnup depth of spent fuel veneers, utilizing transmitted thermal neutron imaging technology. We have significantly enhanced the design of a thermal neutron moderator collimator, leading to remarkable improvements in the quality of the thermal neutron beam. Following moderation by the collimator, the ultimate thermal neutron injection rate at the designated sample location exceeds 103 n/cm2, with thermal neutrons comprising over 74% of the collimated neutron beam. This advanced measurement system enables us to obtain a detailed two-dimensional distribution map of thermal neutrons transmitted through spent fuel boards with varying burnup depths. By analyzing the grayscale intensity patterns in these maps, we can accurately evaluate the burnup degree within the simulated spent fuel plate. Furthermore, we establish a correlation between the transmitted thermal neutron count in the imaging field and the burnup depth of the spent fuel veneer. This allows for precise determination of burnup depth through the analysis of the two-dimensional distribution of transmission thermal neutron intensity. Our findings demonstrate the feasibility of a scheme for detecting the burnup depth in spent fuel boards based on transmission thermal neutron imaging technology, and obtained a linear relationship between the neutron transmission count and the burnup depth of the spent fuel plate, laying a solid theoretical foundation for future research and development of testing equipment to assess burnup in spent fuel veneers.

Observed Correlation Between Local Topography and Passive Neutron Measurements From the Dynamic Albedo of Neutrons (DAN) Instrument on the Mars Science Laboratory (MSL) Rover

AbstractThe Dynamic Albedo of Neutrons (DAN) instrument on the Mars Science Laboratory Curiosity rover primarily measures neutrons that have undergone interactions with rocks and materials in the rover's local environment. As the rover ascends Aeolis Mons, it may encounter more extreme local topography (e.g., cliffs, gullies, canyons). We present three parts of the rover's traverse in which local topography, expressed as the average local relief relative to the rover, is moderately to strongly correlated with an increase in passive thermal neutron count rates. These increases in count rates are consistent with results from radiation transport models of the instrument's performance near simulated topographic features. Additional DAN measurements in areas of high average local relief (>0.25 m) within 5 m of the instrument could bolster this correlation. DAN's sensitivity to topography in its passive mode could be utilized as a new measurement capability and has implications for the operation of future landed missions carrying neutron spectrometers (e.g., VIPER, MoonRanger, Lunar‐VISE).

Design of a compact and portable space neutron spectrometer based on Monte Carlo

With the rapid development of space exploration, the detection of space neutron radiation is becoming increasingly important. The currently widely used Bonner sphere spectrometer have drawbacks such as large size and weight, as well as low fault tolerance, when detecting space neutron spectra. This paper describes in detail a new type of space neutron spectrometer (SNS), which has two different specifications to adapt to the directional and non-directional neutron field environment, and can measure the directional neutron energy spectrum. For the directed neutron field, SNS integrates 12 3He thermal neutron counters (diameter 3 cm: 3, diameter 4 cm: 6, diameter 5 cm: 3) and uses cylindrical polyethylene as a moderator. For non-directed neutron fields, SNS integrates 9 3He thermal neutron counters (diameter 3 cm: 4, diameter 4 cm: 3, diameter 5 cm: 2) located in a single structure made of polyethylene, boron-containing polyethylene and gadolinium. The device is capable of providing a strong directional response in the energy range of thermal neutrons up to 20 MeV, with little sensitivity to neutrons coming from directions other than the axis of the cylinder. The Monte Carlo transport code FLUKA was used to determine the final configuration of the instrument, including the arrangement, number, and position of thermal neutron counters. In addition, the response matrix of the instrument was calculated using FLUKA code. This device can replace traditional Bonner sphere spectrometer for measuring space neutrons, and it also provides reference value for downsized and lightweight neutron spectrometers on the ground.

Design and performance evaluation of a compact thermal and fast neutron spectrometer

Measurement of the space neutron radiation environment has put forward new requirements for the miniaturization and lightweight design of neutron spectrometers. In this paper, a compact broad-spectrum neutron spectrometer was designed, in which the fast neutron detector was based on the EJ-254 scintillator-coupled with SiPMs array, and the thermal and epithermal neutron detectors were based on the NaI(Tl + Li) (NaIL) scintillator coupled with SiPMs array encapsulated in Al and Cd shells, respectively. The overall weight of the neutron spectrometer was only 2.4 kg, and its dimensions were 154 mm × 123 mm × 115 mm. The fast neutron energy spectrum was obtained based on the capture-gated method and the Maximum Likelihood Expectation Maximization (MLEM) unfolding algorithm; the counts of thermal and epithermal neutrons were obtained based on the optimized frequency-domain Pulse Shape Discrimination (PSD) method and cadmium difference method. The proposed neutron spectrometer with all applied methods was tested as a whole system with a252Cf and a241Am–Be source. The spectrum result was compared to an ISO standard source. The accuracy of fast part of 0.5–10 MeV was 5.9% in definition of Relative Root Mean Squared Error (RRMSE). The thermal and epithermal neutron counts were far higher than standard spectrum, and we proved it being effected by laboratory introduced neutron scattering through a simulation, thus the thermal and epithermal flux measurements was believed to be decent. The Figure of Merit (FoM) of the NaIL's PSD neutron gamma discrimination were 2.57 and 2.90 respectively.

Effects of soil moisture variations on the neutron spectra measured above ground: feasibility of a soil moisture monitor system based on neutron moderating cylinders

Mapping the soil moisture is a key activity in water management and sustainable agriculture, especially in regions characterised by fragile agri-food systems and water scarcity. Cosmic Ray Neutron Sensors (CRNS) is a contactless nuclear technology used for estimating soil moisture (SM) content on a 20–30 m scale at the landscape level. Very interestingly, this corresponds to the so-called intermediate scale gap between the local probes, operating on the meter scale, and the satellite-based technologies, working on the kilometre scale and above. In state-of-art CRNS, the cosmic neutrons degraded by the soil are simply counted by a slightly moderated thermal neutron counter. After a calibration procedure, the SM is inferred by combining this count rate with environmental parameters: the atmospheric pressure, temperature and the air humidity. As the SM affects not only the environmental neutron fluence rate but also its energy distribution, this study was organised in such a way to understand if a CRNS with spectrometric capabilities could provide improved information on the SM distribution. To this aim, an environmental neutron spectrometer was designed by extending the Bonner Spheres to a more sensitive system made of moderating cylinders embedding long BF3\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$ _3 $$\\end{document} proportional counters, the Moderating Cylinders Spectrometer (MCS). Relying on literature environmental neutron spectra, corresponding to different SM values in a standardised soil, the count rates in the MCS were calculated for different values of SM. To simulate various counting scenarios, these count rates were associated to different levels of “realistic” uncertainties and unfolded by means of the FRUIT code. The resulting neutron spectra are compared to the literature ones, allowing at estimating the resolving power of the spectrometer in terms of SM.

Simulation study of the lunar water ice detection based on a small neutron generator

Lunar water ice is an important resource to support the construction of lunar base and other deep space activities, and it is crucial to detect lunar water ice. In this paper, we studied the detection of lunar water ice based on a small neutron generator through Monte-Carlo simulation. The schematic of the detection system and the principle are explained. The simulation results demonstrate that the change of thermal neutron flux with the increase of water content is most obvious and clear. There is a positive correlation between thermal neutron counting and water content in lunar soil. Using thermal neutron counting, the water content can be obtained by inversion.

Experimental Study of the influence of PVC casing thickness and well fluid density on prompt fission neutron uranium logging and its corrections

Prompt fission neutron uranium logging is a direct method for measuring uranium content.This method is not influenced by uranium-radium disequilibrium, and provides an accurate and direct way to obtain the uranium content of the formation. However, in field logging, the results of logging are susceptible to the influence of lithology, well fluid and casing. Thus, it must be corrected. Based on the“epithermal /thermal neutron counting rate ratio”(i.e., NE/T ) theory of uranium quantification, a uranium block model is built to simulate field production boreholes. Meanwhile, the influences of PVC casing thickness, well fluid composition and well fluid density on the decay rate of epithermal neutrons and thermal neutrons and the error correction method are investigated. The experimental results show that NE/T increases linearly with increasing thickness of the PVC casing, and the correction equation is μ PVC = -0.0501 dPVC + 1, with a goodness of fit R 2 = 0.9931. The effect of different well fluid compositions and densities on the decay rate of double neutrons is kaolinite ≥ montmorillonite < chlorite < illite. μ_mud follows a quadratic decay law with increasing well fluid density and the fit R 2 < 0.95. Finally, theaverage corrected uranium quantification error is within 5% of the actual uranium content.

An integrated density correction method of four-detector density logging in cased holes

Density logging is critical to shale oil and gas exploration. In density logging of cased holes, the inelastic gamma count and hydrogen index are closely related to the correction of formation density. Aimed at the problems of obtaining the pure inelastic gamma count and correcting the hydrogen index in density logging, the relevant research is implemented in the four detector logging tool. Firstly, a technique to separate the pure inelastic gamma count for density logging is provided. It is based on the pure count coefficient. Then on the basis of determining the density through the pure inelastic count ratio, the hydrogen index is corrected by the epithermal neutron and thermal neutron count ratio. Finally, the integrated density correction equation is proposed. The accuracy of this method is checked using a Monte Carlo model, which provides a strong experimental foundation. By examining the influencing factors, the results show that while the density after Halliburton correction is significantly impacted by borehole size, lithology, salinity, and borehole fluid, the density after integrated correction is least affected by these variables. Besides, the integrated density correction method is more favorable for large borehole size, low salinity and sandstone formation. Through logging examples, it is discovered that the overall uncertainty of the density log after the integrated density correction is lower than the Halliburton density correction log, with a maximum of only 0.013 g/cm3, indicating that the density results are more reliable. The study's findings offer a correction technique for precisely estimating the density of cased holes.

Proof-of-concept study on a water phantom-based neutron spectrometer: Experimental test with 252Cf and 241Am-Be sources

Boron neutron capture therapy (BNCT) is a promising cancer treatment that uses energetic ions released from 10B(n, α)7Li reactions. Accurate assessment of neutron energy spectra is important for simulation-based evaluation of neutron doses during BNCT. In this study, a proof-of-concept study was conducted for a neutron spectrometry technique that involves the use of a water phantom, which is commonly used for quality assurance in BNCT, as a moderator. The technique involves applying unfolding to the count rate distribution of the thermal neutron counter measured within the phantom to derive the energy spectrum. We performed experiments using a spherical 3He proportional counter in neutron fields generated by 252Cf and 241Am-Be sources. The results demonstrated that the spectrometer reasonably reproduced neutron spectra and showed the potential of using a water phantom as a moderator for such a technique.

Evaluation of neutron attenuation properties using helium-4 scintillation detector for dry cask inspection

In this paper, we demonstrate the neutron attenuation of dry cask shielding materials using the S670e helium-4 detector manufactured by Arktis Radiation Ltd. In particular, two materials expected to be applied to the TN-32 dry cask manufactured by ORANO Korea and KORAD-21 by the Korea Radioactive Waste Agency (KORAD) were utilized. The measured neutron attenuation was compared with our Monte Carlo N-Particle Transport simulation results, and the difference is given as the root mean square (RMS). For the fast neutron case, a rapid decline in neutron counts was observed as a function of increasing material thickness, exhibiting an exponential relationship. The discrepancy between the experimentally acquired data and simulation results for the fast neutron was maintained within a 2.3% RMS. In contrast, the observed thermal neutron count demonstrated an initial rise, attained a maximum value, and exhibited an exponential decline as a function of increasing thickness. In particular, the discrepancy between the measured and simulated peak locations for thermal neutrons displayed an RMS deviation of approximately 17.3–22.4%. Finally, the results suggest that a minimum thickness of 5 cm for Li-6 is necessary to achieve a sufficiently significant cross-section, effectively capturing incoming thermal neutrons within the dry cask.

A new neutron-gamma porosity measurement method for pulsed neutron logging tools

As an alternative method for the compensated neutron porosity measurement method, the neutron-gamma porosity measurement method is widely used in cased holes. However, the neutron-gamma porosity measurement method has some problems, such as the lack of detailed theoretical analysis and low sensitivity in high-porosity formations. The purpose of this paper is to clarify the theory of neutron-gamma porosity measurement, improve the measurement sensitivity, and eliminate the effects of environmental factors. Based on the neutron diffusion theory and the gamma transport theory, the distribution of capture gamma rays is described. A new neutron-gamma porosity measurement method is developed based on inelastic and capture gamma-ray distribution. The applicability of the new method in different environments is analyzed using the Monte Carlo method, and the effectiveness of this method is verified using the synthetic model. The results indicate that the capture gamma-ray ratio is only related to the source-detector distance, gamma-ray attenuation length, and neutron migration length. By combining density, different neutron cross sections, inelastic gamma-ray count ratio, and capture gamma-ray count ratio, the parameter secondary gamma-ray hydrogen index (SGH) related to hydrogen index can be extracted. The dynamic range of SGH is much higher than thermal neutron count ratio and the capture gamma-ray count ratio. The errors in porosity calculation by this method under different borehole and lithologic conditions are generally less than 2 pu. The calculation error of this method is close to that of the compensated neutron porosity measurement method. In addition, the applicability of the new method is analyzed without density data, and the results indicate that the calculation results are basically consistent with the results of calculation with density data. The new method has great potential for extensive applications in cased holes or open holes as an alternative for the compensated neutron porosity logging measurement method.

Design of a compact long counter with an improved response using multiple point-like thermal neutron counters

Design of a compact long counter with an improved response using multiple point-like thermal neutron counters

Correlation between thermal neutrons and soil moisture measured by ENDA

Hadrons are the “skeleton” of extensive air shower (EAS). They possess favorable information concerning composition and energy of cosmic ray. Thermal neutrons generated by the EAS hadrons in the ground as well as charged particles in EAS front plane can be detected by Electron-Neutron detector (EN-detector). A prototype of EN-Detector Array (ENDA), ENDA-16 was built at Large High Altitude Air Shower Observatory (LHAASO) to test its performance of detection of cosmic ray composition and energy spectrum. It has been proved in former work that there is a decrease of thermal neutrons detected in rainy season. For quantitative evaluation of influence of soil moisture on thermal neutrons, at the center of ENDA-16, five soil moisture meters are installed to record soil moisture. Negative correlation between thermal neutron counting rate and soil moisture is obtained. Moreover, it is demonstrated that a soil depth 0.5 m over the soil moisture sensor is enough for monitoring negative correlation between thermal neutron counting rate and soil moisture. The results provide us a method to correct the experimental data during the rainy season so as to reduce systematic uncertainty of thermal neutron measurement in the ENDA experiment.

Investigation of CLYC-6 for thermal neutron detection and CLYC-7 for fast neutron spectrometry

Highly enriched 6Li and 7Li Cs2LiYCl 6 (CLYC) scintillators were investigated in view of their application as thermal neutron counter (CLYC-6) and as fast neutron spectrometer (CLYC-7) in the recently developed B-RAD radiation survey metre, an instrument designed to work in regions of high magnetic fields, currently equipped with a probe for photon dose rate measurements and γ-spectrometry. Both crystals were irradiated with a 2.5 MeV mono-energetic neutron beam from a Deuterium–Deuterium (D–D) generator and characterised in terms of neutron/γ-ray (n/γ) discrimination capability by the Pulse Shape Discrimination (PSD) method, energy resolution and quenching factor. Both crystals exhibit an excellent n/γ discrimination, quantified by a Figure of Merit higher than 2. The measured thermal energy resolution of CLYC-6 is 6%. The quenching factor for the 6Li(n,t)α thermal neutron reaction is 0.65 and for the 35Cl(n,p)35S reaction is 0.91. The measured neutron detection efficiency per unit volume of CLYC-6 is 60% higher than that of a 3He proportional counter. CLYC-7 was tested as a fast neutron spectrometer below 10 MeV, when irradiated with continuum neutron spectra from Am–Be and 252Cf sources. A proton/α-particle (p/α) discrimination was performed exploiting the PSD technique, without any unfolding procedure. The resulting spectra match the corresponding ISO spectra better than the spectra without the p/α discrimination, especially in the range 3 MeV to 5 MeV. A more sensitive PSD algorithm might help to improve particle discrimination at lower energies, while different reaction channels arise at higher energies preventing a spectrometry analysis.

Investigation of the stray neutron fields produced from proton irradiated Ta, Co-59, Ni, Fe, Inconel X750 and stainless steel 310 targets in a materials testing laboratory environment

A nested neutron spectrometer was used to measure the stray neutron fluence and ambient dose equivalent produced by protons from a 4 MV tandem accelerator on various target materials at the Reactor Materials Testing Laboratory (RMTL), Queen's University, Kingston, Ontario. The nested neutron spectrometer utilizes a Helium-3 proportional counter and measures the thermal neutron count rate as a function of moderator shell-thickness. Using custom unfolding software, neutron spectra and neutron dosimetric quantities were derived at various locations within the RMTL. The nested neutron spectrometer can be operated in both a pulse mode and current mode allowing for measurements over a wide range of accelerator beam currents. All measurements conducted at the facility were within the main accelerator chamber and experimental target rooms. Target room measurements were conducted with a variety of target materials representative of those to be used for research planned at the RMTL.

Development of a compensation system for thermal neutron measurements based on Cd(Cu)-covered NaI(Tl) detectors

A compensation system based on NaI (Tl) detectors and cadmium sheet was proposed for thermal neutron detection in the neutron/gamma field. The system consisted of two identical NaI (Tl) detectors covered with copper and cadmium sheets, respectively. The Cd-covered NaI (Tl) recorded the prompt gamma rays produced by the 113Cd (n, γ)114Cd reaction, while the Cu-covered structure “rejected” the background gamma rays. The moderated thermal neutron fluxes at different distances from an Am–Be neutron source, calibrated by the activation foil method, were investigated using the compensation system. The thermal neutron detection efficiency of this compensation system was compared to those obtained by the CdZnTe detector and He-3 proportional counter. The feasibility of the compensation system used as a thermal neutron counter was demonstrated by the obtained results.

Sensitivity of the simulation of passive neutron emission from UF6 cylinders to the uncertainties in both 19F([formula omitted],n) energy spectrum and thick target yield of 234U in UF6

Interest in safeguards verification measurements using passive thermal neutron counting to assay 235U content in large 30B UF6 canisters has grown in recent years. The prohibitively high cost and impracticality of using reference 30B calibration cylinders extensively will likely make accurate simulations of increasing interest. Accuracy of the simulated response will define the confidence in the predicted response and the extent to which simulations can reasonably be relied upon. With 234U driven 19F(α, n) reactions being the main neutron source in low enriched UF6 the uncertainties of the 19F(α, n) energy spectrum and the thick target yield of 234U in UF6 propagate into the uncertainty in the predicted response and represent a major influence of basic nuclear data. Here sensitivity of the simulated total (Singles) and coincidence (Doubles) count rates are assessed for the Passive Neutron Enrichment Meter using six potential 19F(α, n) neutron energy spectra over a range of enrichments and material distributions. The results indicate that variations in the Singles and Doubles due to simulated (α, n) neutron spectrum are less than 1.5% for this set of simulated neutron spectra, with dependence varying inversely with enrichment. Singles uncertainty is only slightly less than that of the thick target 19F(α, n) yield corresponding to the primary neutron source, whereas the 19F(α, n) yield dependence of the Doubles is reduced by the non-negligible 238U(SF) coincident neutron emissions. Based on available thick target 19F(α, n) yield estimates the uncertainty is on the order of 5%, establishing this as the main nuclear data limitation when simulating thermal neutron detectors response for 30B UF6 storage cylinders. Based on these findings, it appears that the measurement and evaluation of the thick target 19F(α, n) yield for uranium hexafluoride is due.

Application analysis on the different neutron gamma density (NGD) logging methods

Neutron gamma density (NGD) logging is the most promising alternative to the traditional density logging (GGD), which is of significance for resolving the radiation and safety issues in oil industry. However, due to the different HI correction methods, multiple NGD methods based on the fast neutron, thermal neutron, and capture gamma detection coexist in the well-logging field, and show considerable differences in the tool specifications. To clarify these differences and guide the NGD development, three typical NGD methods using the fast neutron count ratio, thermal neutron count ratio, and capture gamma count ratio (abbreviated as NGD-FC, NGD-TC, and NGD-CC methods) are selected as representatives for comparative study. Using the Monte Carlo simulation, an integrated NGD tool model was established for studying the differences of three NGD methods in the logging responses, data processing methods, and environmental applications. Research shows that, although the three NGD methods have different measurement systems and data processing methods, the three methods can get rid of the HI effect and obtain accurate formation density. The changes of wellbore size and wellbore fluid have similar and significant impact on the three NGD methods and lead to large density errors, especially for the large-size wellbore or wellbore gas conditions. In the different lithology conditions, three methods have good performances, but the NGD-FC and NGD-CC methods have smaller density errors than NGD-TC method. Compared to the other two NGD methods, the NGD-FC method also has a perfect performance in the oil or gas-saturated formation, while NGD-TC and NGD-CC methods have extremely large errors in the gas-saturated formation. Besides, the NGD-FC method are hardly affected by the formation water salinity, the NGD-TC method is slightly affected, while the NGD-CC method is greatly affected. This study can provide a guidance for the tool design, data processing and environment correction of the NGD technology.

Thermal Neutron Cross Section Logging based on compensated neutron logging

In this paper, the compensated neutron logging is transformed into the Thermal Neutron Cross Section Logging (TNXS) to improve the application effect. Because the compensated neutron logging measures the neutron count, TNXS cannot be directly measured by the compensated neutron logging tool. Moreover, the thermal neutron ratio cannot be directly converted into the TNXS due to the influence of lithology and fluid. The ratio of thermal neutron count is related not only to the formation hydrogen index, but also to the formation density and the formation capture cross section. The density and the capture cross section of formations can be used to reduce the influence of lithology and fluid. Therefore, the thermal neutron counting ratio can be converted into the TNXS by density and the capture cross section of formations. The accuracy of the porosity calculated by the TNXS is studied based on the Monte Carlo method. The results show that the thermal neutron counting ratio can be accurately converted to the TNXS by density and the capture cross section of formations. The porosity calculated by TNXS is closer to the formation porosity than that calculated by the compensated neutron logging, especially in gas-bearing formations and complex lithologic formations.

Relative luminosity measurement with Timepix3 in ATLAS

The capability of Timepix3 detectors installed in ATLAS to measure luminosity is evaluated. It is described how noisy pixels are identified and excluded. Two different methods for luminosity determination, i.e. cluster counting and thermal neutron counting are described and compared with each other. The achieved short-term relative precision with both methods is determined by modeling the luminosity curve. It is shown that using cluster counting a short-term relative precision of < 0.5 % can be achieved for 60 s time intervals. For thermal neutrons, a short-term relative precision (for 60 s intervals) of ≈ 2 % was found. Hereby statistics was the limiting factor. The findings are discussed in view of Timepix3 upgrade plans for LHC Run-3.