Neutron Transmission Research Articles

Development of B and BN thin films for in situ neutron beam monitoring

A B-based thin-film neutron monitor is developed for the in situ neutron beam monitoring of the Korea Atomic Energy Research Institute neutron depth profiling system. Thin-film samples are prepared via radio-frequency (RF) sputtering. The RF power, Ar partial pressure, and deposition temperature are varied to optimize the growth and thickness uniformity of the samples. The neutron transmission ratio and collision heating of the designed samples are estimated via Monte Carlo simulation. The neutron transmission ratios of B and BN films prepared in this study are 92.46% and 93.93%, respectively, and the effect of temperature increase is negligible. The displacement per atom (DPA) rates of the B and BN thin films are lower than those of other known B neutron flux monitors, based on the results of DPA calculation for estimating the defect of the sample via neutron irradiation.

Characterization of Deformation by Cold Rolling in Ferritic Steel Containing Cu Particles Using Neutron Transmission Analysis

Characterization of Deformation by Cold Rolling in Ferritic Steel Containing Cu Particles Using Neutron Transmission Analysis

Quantifying water in spent fuel assemblies with neutrons: A simulation-based approach

A simulation-based study concerning three non-destructive approaches by which water in spent nuclear fuel assemblies might be quantified with neutrons is described. Three fuel types have been considered: thirty-six spent Advanced Gas-cooled Reactor (AGR) fuel pins contained in a stainless-steel can; a prototype fast reactor (PFR) spent fuel assembly and a light water reactor (LWR) spent fuel assembly – with the PFR and LWR assemblies containing mixed oxide fuels. The three approaches are investigated with MCNP5 based on neutron interrogation: fast neutron detection of the perturbed fast neutron flux and two approaches based on the 2.223 MeV γ-ray photon emission resulting from the 1H(n,γ)2H capture reaction, stimulated with moderated and unmoderated neutrons. For each of these approaches the following perspectives have been considered: neutron transmission, the influence of radiation emitted by the associated spent fuel inventory, and the dependencies on water quantity and location. The moderated γ-ray assay technique is observed to return the most significant distinction between scenarios with and without water in the fuel assemblies, notwithstanding the need for efficiency response functions to be overlaid on the neutron models as part of the interpretation of the MCNP model results. Water present in all of the spent fuels considered is observed to perturb the incident neutron flux to a degree significant within the simulation uncertainties. The AGR and PFR cases are not undermined by the intrinsic neutron field from the fuel; whereas for the LWR-MOX case, the simulations suggest that this will limit its potential. Successful detection of water quantities down to 10 g is anticipated for AGR, whereas reduced mass sensitivity is forecast for PFR and LWR-MOX, where it is assumed that the water will be more dispersed. Water located in fewer locations is observed to be harder to discern, suggesting that small, localised quantities of water could pose a challenge for these techniques.

New capability for indirect neutron capture measurements: The DICER instrument at LANSCE

The Device for Indirect Capture on Radionuclides (DICER) implements a new indirect technique for (n,γ) studies in which the neutron capture rate is determined from analysis of resonance neutron transmission data. The DICER instrument and associated radionuclide production at the Isotope Production Facility (IPF), both at the Los Alamos Neutron Science Center (LANSCE), as well radioactive sample fabrication, have been under development in the last few years. First measurements on a radioactive sample (88Zr, t½=83.4 days), which was recently reported to have an extremely large thermal neutron capture cross section and resonance integral [1, 2], are planned for the winter of 2021. A performance overview, brief details on the 88Zr fabrication and proof of good operation results will be presented.

Strain Imaging of Corroded Steel Fasteners Using Neutron Transmission Imaging

Strain Imaging of Corroded Steel Fasteners Using Neutron Transmission Imaging

Elemental analysis of concrete via fast neutron transmission and scattering spectrometry

We report on the development of neutron-based techniques to non-destructively measure the composition of concrete. Previous experimental studies demonstrated the viability of the unfolding technique to determine the ratio of water, sand and cement in well-characterised concrete samples from the transmitted neutron energy spectrum. In this work, we used MCNP6 simulations to demonstrate the extension of the technique to determine elemental compositions from transmitted, or scattered neutron energy spectra. In both cases, the simulated energy spectra provided a reliable method to unfold the composition of samples with known elemental ratios. The precision of the technique was limited by the statistical uncertainties of the simulated spectra, particularly for the case of scattered neutrons. The accuracy of the technique was heavily dependent on the uniqueness of each of the elemental responses, and reasonable prior knowledge of the composition. Given the promising results at this stage, future developments will include the addition of further elements to the response matrix, and experimental verification.

Radiation shielding properties of siderurgical aggregate concrete

The world is changing, and consequently so are policies on the use of natural resources. One of the most convenient ways to reduce the consumption of natural aggregates in the production of more sustainable construction materials is the use of recovered industrial by-products. In this study, concretes are designed using siderurgical aggregates from electric arc furnaces, taking advantage of their high density to use them as radiation shielding concrete. To verify the suitability of these aggregates, four concrete mixes were designed with different aggregates: limestone, siderurgical magnetite aggregates (the most commonly used in the nuclear field). The comparison of the different mixes was carried out focusing on the physical–mechanical properties in the field of ionizing radiation shielding (gamma radiation and neutron shielding) by means of simulations. In addition, an analysis was performed to establish how the w/c ratio and the amount of CEM affect shielding properties. In terms of linear attenuation coefficient and neutron transmission rate, the concrete with siderurgical aggregates shows intermediate capability in comparison with the limestone aggregate and magnetite concrete. The increase in the amount of cement and the w/c ratio caused a decrease in the linear attenuation coefficient and a reduction in the neutron transmission rate, but the variation in the w/c ratio did not have a significant impact on the neutron transmission rate.

New capability for neutron transmission measurements at LANSCE: The DICER instrument

A new capability for neutron transmission measurements has been developed at the Los Alamos Neutron Science Center (LANSCE). The DICER (Device for Indirect Capture Experiments on Radionuclides) instrument is designed to tightly constrain neutron-capture cross sections for short-lived radionuclides that are beyond the reach of direct techniques, via measurement and analysis of neutron transmission resonances on these same radionuclides. The innovative DICER apparatus enables simultaneous sample-in and sample-out measurements, thus minimizing the amount of sample material needed and reducing the measurements time by half. A description of the current status of the new transmission station, along with a performance overview and results from known reactions, will be presented.

Feasibility of gadolinium oxide paint as neutron shielding

A robust Gd2O3 based thermal neutron shielding paint has been produced to act as the neutron attenuation medium to augment radiation shielding and to produce simple low resolution neutron imaging resolution standards. The neutron attenuation of this paint has been studied using a neutron radiographical imaging beamline at the McMaster Nuclear Reactor, and the McMaster Alignment Diffractometer neutron beam instrument. The data indicates a thermal neutron transmission factor of 0.457±0.002 and a mass attenuation coefficient of 89±2gcm2 for a single coat of paint.

A comparative study of promising filter materials for neutron imaging facilities

The effect of sapphire and bismuth single-crystal filters and their combinations on the quality of neutron radiographic images and neutron tomography data has been studied. The parameters of the contrast of the neutron image were analyzed depending on the monocrystalline filters. Neutron transmission spectra were obtained for sapphire and bismuth single crystals. Additionally, the effect of filters on the overall intensity of the thermal neutron beam and the background of gamma-rays was investigated. Based on the obtained data, we assume that a single-crystal sapphire filter can be most effectively used for radiographic and tomographic installations using thermal neutrons.

High Spatial Resolution Neutron Transmission Imaging Using a Superconducting Two-Dimensional Detector

Neutron imaging is one of the most powerful tools for nondestructive inspection owing to the unique characteristics of neutron beams, such as high permeability for many heavy metals, high sensitivity for certain light elements, and isotope selectivity owing to a specific nuclear reaction between an isotope and neutrons. In this study, we employed a superconducting detector, current-biased kinetic-inductance detector (CB-KID) for neutron imaging using a pulsed neutron source. We employed the delay-line method, and high spatial resolution imaging with only four reading channels was achieved. We also performed wavelength-resolved neutron imaging by the time-of-flight method for the pulsed neutron source. We obtained the neutron transmission images of a Gd–Al alloy sample, inside which single crystals of GdAl <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex-math notation="LaTeX">$_3$</tex-math></inline-formula> were grown, using the delay-line CB-KID. Single crystals were well imaged, in both shapes and distributions, throughout the Al–Gd alloy. We identified Gd nuclei via neutron transmissions that exhibited characteristic suppression above the neutron wavelength of 0.03 nm. In addition, the <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex-math notation="LaTeX">$^{155}$</tex-math></inline-formula> Gd resonance dip, a dip structure of the transmission caused by the nuclear reaction between an isotope and neutrons, was observed even when the number of events was summed over a limited area of 15 <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex-math notation="LaTeX">$\mu$</tex-math></inline-formula> m × 12 <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex-math notation="LaTeX">$\mu$</tex-math></inline-formula> m. Gd selective imaging was performed using the resonance dip of <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"><tex-math notation="LaTeX">$^{155}$</tex-math></inline-formula> Gd, and it showed clear Gd distribution even with a limited neutron wavelength range of 1 pm.

Time-Resolved Neutron Bragg-Edge Imaging: A Case Study by Observing Martensitic Phase Formation in Low Temperature Transformation (LTT) Steel during GTAW

Polychromatic and wavelength-selective neutron transmission radiography were applied during bead-on-plate welding on 5 mm thick sheets on the face side of martensitic low transformation temperature (LTT) steel plates using gas tungsten arc welding (GTAW). The in situ visualization of austenitization upon welding and subsequent α’-martensite formation during cooling could be achieved with a temporal resolution of 2 s for monochromatic imaging using a single neutron wavelength and of 0.5 s for polychromatic imaging using the full spectrum of the beam (white beam). The spatial resolution achieved in the experiments was approximately 200 µm. The transmitted monochromatic neutron beam intensity at a wavelength of λ = 0.395 nm was significantly reduced during cooling below the martensitic start temperature Ms since the emerging martensitic phase has a ~10% higher attenuation coefficient than the austenitic phase. Neutron imaging was significantly influenced by coherent neutron scattering caused by the thermal motion of the crystal lattice (Debye–Waller factor), resulting in a reduction in the neutron transmission by approx. 15% for monochromatic and by approx. 4% for polychromatic imaging.

Neutron Tomography of Spent Fuel Casks

Dry casks for spent nuclear fuel (SNF) ensure the safe storage of SNF and provide radiation shielding. However, the presence of the thick casks encompassing several layers of steel and concrete makes inspection of the SNF a challenging task. Fast neutron interrogation is a viable method for the nondestructive assay of dry storage casks. In this study, we performed a Monte Carlo simulation-based study associated with a machine-learning-based image reconstruction method to verify the content of SNF dry storage casks. We studied the use of neutron transmission and back-scattered measurements to assess the potential damage to fuel assemblies or fuel pin diversion during transportation of dry casks. We used Geant4 to model a realistic HI-STAR 100 cask, MPC-68 canister and basket, and GE-14 fuel assembly irradiated by a D-T neutron generator. Several bundle diversion scenarios were simulated. The angular distribution of the neutrons scattered by the cask was used to identify the diversions inside the fuel cask. A fuel bundle with at least 75% of its pins removed can be identified with a drop in the back-scattered signature larger than $$2\sigma$$ compared with a fully loaded scenario. We combined an iterative reconstruction algorithm with a convolutional neural network (CNN) to obtain a cross-sectional image of the fuel inside the cask. The proposed imaging approach allows locating the position of a missing fuel bundle with at least 75% of the pins removed when performing tomographic imaging of a canister with an overall scan time of less than two hours, when using a commercial neutron generator with a source strength of $$10^{10}$$ n/s in the 4 $$\pi$$ solid angle.

Influence of diffusion and viscosity on the field asymmetry of the effect of gravity in the critical fluid

A brief review of the results of studies of the effect of gravity in inhomogeneous substance near a critical state of a critical fluid (CF) has been presented in paper, based on the data of light scattering, refractometry, and slow neutron transmission methods.Based on these data, the field-altitude asymmetry of various properties of an inhomogeneous substance has been analyzed, namely order parameter Dr(z), scattered light intensity I(z), density gradient dr(z)/dz of the substance. It had been shown that the field-altitude asymmetries of the scattered light intensity I(z)~dr/dm(h) and the density gradient dr(z)/dz~dr/dh(h) of the substance are diametrically opposite. The different altitudinal asymmetry of these quantities dr/dh(h) and dr/dm(h) is explained in paper by the altitude asymmetry of the derivative of the chemical potential dm/dh, and hence with the altitude asymmetry of the chemical potential Dm(h)&gt;&gt;h in the external field h.To the present time, the physical mechanism of the altitude asymmetry of the gravity effect has not been studied. In this regard the mechanism of the formation of the vertical asymmetry of the internal critical field Dm(h) has proposed in paper to be associated with the kinetic characteristics of the inhomogeneous critical fluid: the diffusion coefficients D (h) and viscosity coefficients h(h), when the system passes from a homogeneous state to an inhomogeneous one under the action of an internal asymmetric fields |DU(z)|= |Dm(z)|&gt;&gt;|h=rcgz/Pc|. For this purpose, a high-pressure cell with a height L, with a critical filling density of the substance is considered in paper.It has been shown that when the system is under critical density filling by substance =rc the critical level of substance z = 0 with the critical density rc at the critical temperature Tc is realized above the middle of the sample with an inhomogeneous substance. Based on the literature data of P-V-T-measurements and the gravity effect in benzene and ethane, the values of the altitudinal change in the internal critical field have been found. It has been shown that the value of the critical internal inhomogeneous field in the inhomogeneous critical fluid significantly exceeds the variable of the Earth's gravity |DU(h,Tc)|= |Dm(h,Tc)&gt;&gt;|h| It has been also shown that the magnitude of this field according to the cubic law depends on the critical temperature Tc of the substance: |Dm(z,Tc1)/|Dm(z,Tc2) » (Tc1/Tc2)3.

Synthesis and characterization of B2O3-Ag3PO4-ZnO-Na2O glasses for optical and radiation shielding applications

In this paper, a new glass system based on (55-x) B2O3-x Ag3PO4-45 ZnO-5 Na2O (x = 0.0, 5.0, 10.0 and 15.0 mol%) is synthesized via melt quenching technique and the corresponding glass specimens named as BZN-AP1, BZN-AP2, BZN-AP3, and BZN-AP4. X-ray diffraction (XRD), UV-IR are employed to evaluate the structure of the samples. The XRD analysis shows that all glass samples are amorphous in nature. UV-IR shows that the increase in Ag3PO4 concentrations results in a red shift in the region of lower photon energy and the energy band gap decreases as the concentration of Ag3PO4 increases. The neutron radiation shielding’s characteristics such as neutron coherent, incoherent and absorption cross section and transmission factor NTF of the glass sample are evaluated. The value of the neutron transmission factor NTF reduced as the thickness of the glass specimen increases. The fast and slow neutron shielding ability of the present glass system reduced as the Ag3PO4 content increased. However, the increase in Ag3PO4 content decreases the coherent, incoherent and absorption cross sections. The results presented in this work demonstrate the pivotal role of Ag3PO4 content on the synthesis, structure, optical, and neutron shielding properties of (55-x) B2O3 – x Ag3PO4 – 45 ZnO-5Na2O) glasses.

Design and Characterization of a Neutron-Friendly Lithium-Ion Battery Coin Cell for Extreme Fast-Charging

Lithium-ion batteries (LIBs) have profoundly advanced the development of electric vehicles (EV). However, one of the remaining bottlenecks in their widespread deployment is the long charging time typically required for commercial LIBs. There is a global push towards extreme fast charging (XFC) of EV batteries to reduce their charging times to 10-15 minutes.1 However, XFC causes severe degradation in the electrochemical performance of LIBs, which is mainly attributed to irreversible Li plating on the surface of the graphite anode rather than intercalation of Li ions inside the graphite as the battery is fast-charged.2 Since Li plating has been identified as one of the major XFC degradation mechanisms Bs, understanding its origin and spatial heterogeneity across three dimensions (3D) on porous graphite anodes as a function of time necessitates a non-destructive characterization platform to differentiate between graphite and Li.Neutron-based imaging provides the sensitivity to differentiate graphite from Li due to the large difference in their total neutron cross-sections (Li:C ~ 73:11 barns). However, high energy X-rays enable penetrating the metallic battery components. e.g., Cu, Al. We tested the feasibility of simultaneous neutron and X-ray tomography (NeXT) on the BT-2 imaging beamline at National Institute of Standards and Technology Center for Neutron Research.4 Our spatial resolution for both X-rays and neutrons was ~15-20 μm. For ex-situ Li plating detection, we imaged 1) pristine, and 2) a cycled graphite anode.5 For cycled anode, we disassembled the battery pouch cells in their fully discharged condition and harvested the anode after 450 cycles of 9C charging and C/2 discharging. For data analysis, we used 2D bivariate histogram phase segmentation to differentiate different battery materials such as Cu, graphite, etc. However, low X-ray energy (40keV) combined with the flat sample geometry of the graphite strips made it challenging to remove beam hardening artifacts from copper and, therefore, to segment lithium from graphite. In addition, to determine NeXT feasibility for in-situ imaging, we also characterized a fully assembled, uncycled single-layer battery pouch cell. However, the high neutron attenuation by the hydrogen in EC: EMC (3:7) electrolyte and hydrogenous polymeric materials in the Celgard separator made it particularly difficult to segment graphite anode from the separator and positively identify interfaces.In this presentation, we will outline the design and characterization of a neutron-friendly LIB coin cell for extreme fast-charging. Our design is based on a standard 2032 format coin cell with custom modifications to address the dual criteria of electrochemistry and neutron imaging, i.e., good neutron transmission and relative contrast from a functional LIB at 6-C charging rate. We will present three advantages of coin cell geometry: 1) it provides more uniform attenuation throughout the entire tomography scan thus, reducing artifacts; 2) multiple coin cells can be stacked on top of each other for high throughput battery imaging; 3) it reduces the amount of expensive deuterated materials needed while developing the optimum cell configurations. In addition, we will address the need of a deuterated EC: DMC electrolyte in a fluorinated separator for the neutron-friendly battery design (Figure 1). Since the total neutron cross-section of deuterium and fluorine are on the orders of magnitude lower than that of hydrogen, we anticipate that these modifications will increase neutron transmission through the battery, thus improving signal-to-noise ratio and reducing acquisition time especially at low neutron flux facilities. Secondly, they will result in a cleaner electrolyte-separator-anode interface for the detection of plated lithium because of absence of hydrogen. Finally, we will show how these modifications affect the electrochemical performance of our LIB coin cells to determine the onset of Li plating at 6-C charging rate. i-e., capacity fade vs. number of cycles.

Sensitivity simulation on isotopic fissile measurement using neutron resonances

Uranium and plutonium are required to be accounted in spent fuel head-end and major recovery area in pyro-process for safeguards purpose. The possibility of neutron resonance technique, as a non-destructive analysis, was simulated on isotopic fissile analysis for large scale process. Neutron resonance technique has advantage to distinguish uranium from plutonium directly in mixture. Simulation was performed on U235 and Pu239 assay in spent fuel and for scoping examination of assembly type. The resonance energies were determined for U235 and Pu239. The linearity in the neutron transmission was examined for the selected resonance energies. In addition, the limit for detection was examined by changing sample density, thickness and content for actual application. Several factors were proposed for neutron production and the moderated neutron source was simulated for effective and efficient transmission measurement. From the simulation results, neutron resonance technique is promising to analyze U235 and Pu239 for spent fuel assembly. An accurate fissile assay will contribute to an increased safeguards for the pyro-processing system and international credibility on the reuse of fissile materials in the fuel cycle.

Glass Transition in Rice Pasta as Observed by Combined Neutron Scattering and Time-Domain NMR.

Experimental protocols aiming at the characterisation of glass transition often suffer from ambiguity. The ambition of the present study is to describe the glass transition in a complex, micro heterogeneous system, the dry rice pasta, in a most unambiguous manner, minimising the influence of technique-specific bias. To this end, we apply an unprecedented combination of experimental techniques. Apart from the usually used NMR and DSC, we employ, in a concurrent manner, neutron transmission, diffraction, and Compton scattering. This enables us to investigate the glass transition over a range of spatio-temporal scales that stretches over seven orders of magnitude. The results obtained by neutron diffraction and DSC reveal that dry rice pasta is almost entirely amorphous. Moreover, the glass transition is evidenced by neutron transmission and diffraction data and manifested as a significant decrease of the average sample number density in the temperature range between 40 and 60 °C. At the microscopic level, our NMR, neutron transmission and Compton scattering results provide evidence of changes in the secondary structure of the starch within the dry rice pasta accompanying the glass transition, whereby the long-range order provided by the polymer structure within the starch present in the dry rice pasta is partially lost.

The neutron streaming correction method in the homogenization of Pebble-bed HTGR reflector

In pebble-bed high temperature gas-cooled reactors (HTGRs), the outside graphite reflector generally contains void regions, including the boring channels for control rods, absorber balls and coolant flow. Deterministic reactor physics calculation of HTGR is usually based on regular geometries, which requires pre-homogenization to the materials with void regions. However, the volume-weighted homogenization to the HTGR reflector can greatly overestimate the keff, due to the neutron streaming effect in the reflector channels. This paper proposes a convenient method of correcting the neutron streaming effect in the HTGR reflector, by adjusting the homogenized transport cross sections based on preservation of the neutron transmission probabilities. The proposed method has been implemented in the PANGU code for pebble-bed HTGR physics analysis and fuel cycle simulations. Numerical results suggest that the proposed method can effectively reduce the discrepancy in the homogenization of the HTGR reflector.

Evaluation of CdS doped polyester composites regarding gamma and neutron shielding properties

In this study, CdS doped polyester composites were produced at different weight ratios, and the impact of dopant amount on gamma and neutron shielding characteristics of a new composite material were further evaluated: the composite matrix was doped at 5%, 10%, 15% and 20% CdS weight ratios of the prepared composite. Then, the produced polyester materials have been characterized using a scanning electron microscope (SEM) and X-ray diffraction (XRD) patterns. Following the structure characterization, the prepared composites were tested using one gamma spectrometer system based HPGe detector with help of 22Na, 54Mn, 57Co, 60Co, 133Ba, 137Cs, 152Eu and 241Am radioactive point-isotropic sources in order to perform the evaluation in a wide energy range of gamma. The obtained results were compared to the results of Geant4 and WinXCOM simulation codes. On the last step of the presented work, the neutron removal cross sections of produced polyester composites were calculated using partial density of elements and their cross sections, and then, neutron shielding characteristics of the prepared samples were determined with help of Geant4 simulation toolkit. Both tested and simulated results of gamma shielding evaluation showed good agreement, and, for all considered photon energies, mass attenuation coefficient (μ/ρ) values of CdS (20%) were always found to be greater than the μ/ρ values of other samples (5%, 10%, 15%). The results for neutron shielding showed that neutron transmission was decreasing with the increase of CdS in the material. In short, having higher μ/ρ values and lower neutron transmission shows that increasing the amount of CdS in the sample causes the higher radiation shielding property for both gamma and neutron.