Neutron Transmission Research Articles

Improved limits on n→n′ transformation from the Spallation Neutron Source

Conversions between neutrons n and dark matter candidate sterile neutrons n′ have been proposed as a mechanism for baryon number B violation. In the case that there is a small mass difference Δm between the n and the n′ states, oscillations can be induced by compensating for Δm with a magnetic field. A search for such neutron oscillations was performed at the Spallation Neutron Source by looking for anomalous neutron transmission through a strongly absorbing cadmium wafer inside of a 6.6 T magnet. The approach described here saw no regenerated neutrons above background, which provides an improved limit for neutron–sterile neutron transformations for a range of Δm between 0.1 and 1000 neV. Published by the American Physical Society 2024

Non-invasive characterization of the manufacturing process of a Nuragic bronze statuette: a Neutron Imaging study

The Nuragic civilization (Sardinia, Italy, XVIII–VIII Cen. B.C) developed a flourishing bronze metallurgy. The production of Nuragic bronze figurines from Sardinia represents a rich historical archive that provides key information about the iconography, the metal production and casting techniques, and on the development of metallurgy in the Mediterranean basin. Since the question about their manufacturing method remains without definitive answer, the understanding of the Sardinian bronze metallurgy is essential to determine which manufacturing techniques were employed to produce complex bronze artefacts. In the frame of a wider research project relating to Nuragic bronzes, four artefacts, three anthropomorphic statuettes (a warrior, a priestess, and an offering figure), and one miniature of a basket, were made available by Museo Nazionale Preistorico “L. Pigorini” (Roma, IT). In this work we present the results of the analyses conducted on a bronze figurine depicting an iconic type of Nuragic figure: the Priestess. The analysis was performed using White Beam Neutron Tomography (NT) and Bragg Edge Neutron Transmission (BENT) at the Paul Scherrer Institut (PSI) (Villigen, CH). Neutron techniques are nowadays the only available approach for revealing, non-destructively and with good spatial resolution, the morphological and microstructural properties within the whole volume of solid cast metallic artefacts such as this bronze statuette. This work presents the result of a non-invasive analytical investigation on an archaeological bronze artefact, providing outstanding results: from a quantitative analysis of the composition to an in-depth morphological and microstructural analysis capable of unveiling details on the ancient casting methods of the statuette.

Development of a multi-layer high-efficiency GEM-based neutron detector for spallation sources

Neutron detection is nowadays mostly based on 3He gas detectors, but its shortage and the continuous upgrades of the neutron facilities require new devices to perform experiments with maximum performances. This work presents a new detector based on the Gas Electron Multiplier (GEM) combined with several boron layers. This detector combines the features of GEM technology with the properties of boron as a neutron converter and the device is produced to sustain high neutron fluxes with high detection efficiency. The detector has been characterised at the ISIS Pulsed Neutron and Muon Source (UK). Based on the analysis of our results, the detector has shown a good response to thermal and epithermal neutrons reaching a detection efficiency of 16% at 1.8 Å (25 meV). The good detection efficiency (even increasable with the addition of further boron GEM foils) and the good time resolution, make the detector a unique device for the neutron techniques. In particular, its use can easily be envisaged in techniques involving neutron transmission measurements, that require high fluxes impinging on the detectors, with the added bonus of a 2D-resolved capability due to the padded anode.

A comparative investigation of neutron and gamma radiation interaction properties of zircaloy-2 and zircaloy-4 with consideration of mechanical properties

Abstract This study has established the radiation shielding efficacy of zircaloy-2 and zircaloy-4 over a wide spectrum of energy levels. Using the Monte Carlo N-Particle (MCNP) method, the gamma and neutron transmission factors (TF and nTF) were calculated for various energy levels. Zircaloy-2 demonstrated the highest gamma-ray absorption capacity and the lowest neutron absorption capacity among the investigated alloys. The results indicate that zircaloy-2 and zircaloy-4 have nearly the same neutron transmission characteristics. Although many studies have examined the structure and physical characteristics of these materials, there has been a lack of Monte Carlo simulations to comprehensively investigate the correlation between gamma absorption, neutron absorption parameters, and mechanical qualities. This research aims to examine the ability of zirconium and its zircaloy-2 and zircaloy-4 alloys, which are critical materials used in the nuclear industry, to absorb gamma and neutron radiation over a broad spectrum of frequencies. According to the results, zircaloy-2 has the best ability to absorb secondary gamma rays and the highest level of resistance to them. Despite the minimal disparity in the nTF between the two alloys, simulation results have shown that zircaloy-2 has a higher level of neutron transmittance. These results have the potential to expedite the development of novel materials with enhanced attributes for various applications.

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.

Synthesis, photoluminescence, Judd-Ofelt parameters, and high energy neutron/charged particle transmission efficiencies of Nd3+ ion-activated sodium-borate glasses

Sodium borate glasses doped with Nd3+ ions and having the molar formulations 75B2O3 – 15Na2O – 9.5x – 0.5Nd2O3, where x = Bi2O3; SrO and Li2O were prepared by employing the melt quenching technique. The glasses were coded as BiNd, SrNd, and LiNd for x = Bi2O3; SrO and Li2O, respectively. The prepared glasses were characterized for their physical and optical attributes. The influence of the glass composition on the photoluminescence and fast neutron (FN) and charged radiation (CR) transmission efficiencies were also investigated. The density of BiNd, SrNd, and LiNd is 3.32, 2.43, and 3.32 g/cm3, respectively. Optical constants of the glasses showed variations with glass composition. The FTIR study of the glass also revealed the structure of the glasses. The BiNd glass has the most intense PL of all the samples, followed by SrNd glass, while LiNd had the weakest PL emission. The value of for BiNd, SiNd, and LiNd is 0.1029 cm-1, 0.1009 cm-1, and 0.0987 cm-1, respectively. The projected ranges of electrons, protons, helium ions, and carbon ions were generally lower for denser XNd glass. The XNd glasses are potentially useful of optical, shielding and dosimetry applications.

Non-invasive detection of hazardous materials with a thermal-to-epithermal neutron station: a feasibility study towards practical application

The growing scale of the devastation that even a single terrorist attack can cause requires more effective methods for the detection of hazardous materials. In particular, there are no solutions for effectively monitoring threats at sea, both for the off-shore infrastructure and ports. Currently, state-of-the-art detection methods determine the density distribution and the shapes of tested subjects but only allow for a limited degree of substance identification. This work aims to present a feasibility study of the possible usage of several methods available on the thermal-to-epithermal neutron station, VESUVIO, at the ISIS neutron and muon spallation source, UK, for the detection of hazardous materials. To this end, we present the results of a series of experiments performed concurrently employing neutron transmission and Compton scattering using melamine, a commonly used explosive surrogate, in order to determine its signal characteristics and limits of detection and quantitation. The experiments are supported by first-principles modelling, providing detailed scrutiny of the material structure and the nuclear dynamics behind the neutron scattering observables.

Multicomponent intermetallic compound CoCrFeNiSi: Crystal structure and magnetic properties

A novel intermetallic compound with a cubic crystal structure was successfully synthesized within the quinary system of CoCrFeNiSi, boasting an almost equiatomic chemical composition. The crystal structures of the intermetallics and the minor impurity phase has been meticulously investigated using X-ray diffraction (XRD), time-of-flight neutron diffraction and transmission electron microscopy. Magnetic measurements of this high-entropy intermetallic compound revealed that CoCrFeNiSi exhibits ferrimagnetic properties at room temperature. The structural disorder was investigated through theoretical calculations and magnetic moment measurements.

Application of Energy‐Resolving Neutron Imaging to Major‐Component Analyses of Materials Using Four‐Channel Superconducting Detector

We proposed a current‐biased kinetic inductance detector (CB‐KID) as a novel superconducting detector to construct a neutron transmission imager. The characteristics of a superconducting neutron detector have been systematically studied to improve a spatial resolution down to 10 in transmission imaging. In this study, we report the application of the energy‐resolving neutron imaging to investigate major components of materials by analyzing neutron transmission spectra from 1 meV to 500 keV. We succeeded in identifying that copper (Cu) and iron (Fe) are major components respectively in commercial nuts and screws as test samples with the aid of Rietveld imaging of transmission spectra (RITS) program in analyzing transmission spectra in longer wavelengths. The Ti screw was also confirmed by comparing the nuclear resonance absorption measurements and simulations in high‐energy regions. We demonstrated that our superconducting neutron detector is applicable to reveal the transmission spectra in the wide range from cold‐neutron energies to higher neutron energies even up to 500 keV. By selecting distinctive energy regions of pulsed neutrons, we succeeded in mapping the distribution of SmSn3 compound using the strong neutron absorption in samarium (Sm) and the selective nuclear‐resonance dips in Sm. By taking advantage of using CB‐KID in conducting neutron imaging, the CB‐KID method is extensively useful for various purposes in material sciences through energy‐selective neutron spectroscopy from 1 meV to 500 keV. © 2024 Institute of Electrical Engineers of Japan and Wiley Periodicals LLC.

Direct inversion of the Longitudinal ray transform for 2D residual elastic strain fields

We examine the problem of Bragg-edge elastic strain tomography from energy resolved neutron transmission imaging. A new approach is developed for two-dimensional plane-stress and plane-strain systems whereby elastic strain can be reconstructed from its Longitudinal ray transform (LRT) as two parts of a Helmholtz decomposition based on the concept of an Airy stress potential. The solenoidal component of this decomposition is reconstructed using an inversion formula based on a tensor filtered back projection (FBP) algorithm whereas the potential part can be recovered using either Hooke’s law or a finite element model of the elastic system. The technique is demonstrated for two-dimensional plane-stress systems in both simulation, and on real experimental data. We also demonstrate that application of the standard scalar FBP algorithm to the LRT in these systems recovers the trace of the solenoidal component of strain and we provide physical meaning for this quantity in the case of 2D plane-stress and plane-strain systems.

High-Resolution Modeling Without Computation Slowdown for PETALE in CROCUS

In a collaboration between Ecole Polytechnique Fédérale de Lausanne (EPFL) and CEA, in the fall of 2020, the experimental Programme d’Étude en Transmission de l’Acier Lourd et ses Eléments (PETALE) was successfully carried out in the CROCUS reactor of EPFL. This article presents and compares the methods tested in the modeling of the experiments, specifically focusing on the metal reflectors installed at the periphery of CROCUS. A basic design model consisting of a few cuboids was refined to a fully detailed version, without impacting the run time of simulations. Notably, each reflector sheet of PETALE was segmented into 121 voxels based on topological measurements. This detailed voxelization did not affect calculation times, thanks to the use of three-dimensional lattices as available in Serpent 2. Profiling of the simulations revealed the high computational surface transformations associated with Serpent 2 and highlighted the efficiency benefits of factorizing these into universe transformations. As the CROCUS simulations were carried out using a modified build of Serpent 2, additional simulations were also performed using a standard version of Serpent 2 with a GODIVA model as a neutron source to ensure that the findings are generalizable. These additional tests confirmed the initial results, with significant performance variations observed between the models, particularly larger in surface-tracking mode than in delta-tracking mode. Consequently, the modeling method may therefore be applied to future high-fidelity modeling of neutron transmission and shielding experiments.

Development of novel composite materials containing rice bran wax and waste polyethylene for neutron shielding applications

Neutrons are utilized in diverse fields, such as physics, biology, chemistry, geology, agriculture, medicine, mining, military, and space research. Neutrons interacting with human tissues and cells can cause harm to soft tissues, such as the cornea, and alter cell functionality, leading to the cessation of cell proliferation and resulting in cell/tisssue damage or cancers. Effective protective materials are necessary to enable the safe use of neutrons in these applications. In this study, several samples of novel types of composites were prepared by blending rice bran wax (RBX) with recycled polyethylene (rPE) at five different ratios ranging from 0% to 50%. The RBX in the composites enhances their neutron absorption capacities and while the rPE in the composites reduce their costs. Several essential fast neutron shielding parameters, such as half-value layer, effective removal cross-section, mean free path, and neutron transmission ratio, were calculated with Monte Carlo simulation GEANT4 code. Absorbed dose amount by the samples were determined using a241Am-9Be neutron source. The theoretical and experimental values of the shielding capacities of the novel composites were compared with those of paraffin and traditional concrete. Compared with the reference samples, the new RBX composite samples had higher neutron radiation absorption capacities. The functional groups of the RBX and rPE were identified using Fourier transform infrared (FTIR) spectrophotometry (Bruker Vertex 70v FTIR spectrometer with the attenuated total reflection (ATR) module) in the spectral area of 400–4000 cm−1 with 32 scans. The thermal chareteristics of the samples were determined with the differential scanning calorimeter Mettler-Toledo/DSC1/700. The developed composite materials are useable as protective material against neutron radiation in radiation settings, such as radiotherapy rooms, within and outside nuclear power plants, as well as in settings involving the storage or transportation of radioactive waste.

Neutron transmission imaging system with a superconducting kinetic inductance detector

Abstract We optimized the design and operating conditions of our superconducting neutron detector to improve spatial resolution. We obtained the best spatial resolution of 10 μm when a Gd Siemens star pattern was mounted in close contact with the detector. Although there is a trade-off between a spatial resolution and an easiness of replacing samples, we built our superconducting neutron imaging system for measuring in both the room-temperature samples with a proper collimation ratio L/D for achieving a reasonable spatial resolution and a cryogenic temperature with the best spatial resolution for certain purposes. In this study, we obtained neutron transmission images of various samples when they were cooled down with the superconducting neutron detector. We compared the effect of a different sample-detector distance on a spatial resolution when the samples were placed either at cryogenic temperature or at room temperature. We also confirmed that our CB-KID sensor was able to observe the neutron transmission coefficient over wider energies of pulsed neutrons. We found the appearance of clear Bragg dips by the measurements of natural FeS2 single crystals and succeeded in mapping the distribution of differently-oriented crystals by choosing several Bragg dips of the FeS2 crystals to compose the transmission images.

A re-activation model for 169Yb intensity modulated brachytherapy sources accounting for spatiotemporal isotopic composition.

Intensity modulated brachytherapy based on partially shielded intracavitary and interstitial applicators is possible with a cost-effective 169Yb production method. 169Yb is a traditionally expensive isotope suitable for this purpose, with an average γ-ray energy of 93keV. Re-activating a single 169Yb source multiple times in a nuclear reactor between clinical uses was shown to theoretically reduce cost by approximately 75% relative to conventional single-activation sources. With re-activation, substantial spatiotemporal variation in isotopic source composition is expected between activations via 168Yb burnup and 169Yb decay, resulting in time dependent neutron transmission, precursor usage, and reactor time needed per re-activation. To introduce a generalized model of radioactive source production that accounts for spatiotemporal variation in isotopic source composition to improve the efficiency estimate of the 169Yb production process, with and without re-activation. A time-dependent thermal neutron transport, isotope transmutation, and decay model was developed. Thermal neutron flux within partitioned sub-volumes of a cylindrical active source was calculated by raytracing through the spatiotemporal dependent isotopic composition throughout the source, accounting for thermal neutron attenuation along each ray. The model was benchmarked, generalized, and applied to a variety of active source dimensions with radii ranging from 0.4 to 1.0mm, lengths from 2.5 to 10.5mm, and volumes from 0.31 to 7.85 mm3, at thermal neutron fluxes from 1×1014to1×1015 n cm-2 s-1. The 168Yb-Yb2O3 density was 8.5g cm-3 with 82% 168Yb-enrichment. As an example, a reference re-activatable 169Yb active source (RRS) constructed of 82%-enriched 168Yb-Yb2O3 precursor was modeled, with 0.6mm diameter, 10.5mm length, 3 mm3 volume, 8.5g cm-3 density, and a thermal neutron activation flux of 4 × 1014 neutronscm-2s-1. The average clinical 169Yb activity for a 0.99 versus 0.31 mm3 source dropped from 20.1 to 7.5Ci for a 4×1014ncm-2s-1 activation flux and from 20.9 to 8.7Ci for a 1×1015ncm-2s-1 activation flux. For thermal neutron fluxes ≥2×1014ncm-2s-1, total precursor and reactor time per clinic-year were maximized at a source volume of 0.99 mm3 and reached a near minimum at 3mm3. When the spatiotemporal isotopic composition effect was accounted for, average thermal neutron transmission increased over RRS lifetime from 23.6% to 55.9%. A 28% reduction (42.5 days to 30.6 days) in the reactor time needed per clinic-year for the RRS is predicted relative to a model that does not account for spatiotemporal isotopic composition effects. Accounting for spatiotemporal isotopic composition effects within the RRS results in a 28% reduction in the reactor time per clinic-year relative to the case in which such changes are not accounted for. Smaller volume sources had a disadvantage in that average clinical 169Yb activity decreased substantially below 20Ci for source volumes under 1mm3. Increasing source volume above 3 mm3 adds little value in precursor and reactor time savings and has a geometric disadvantage.

Neutron imaging of exchange flow between biopore and matrix for Bt versus C horizons

AbstractThe surface coating of biopores plays a pivotal role in mediating mass exchange between the biopore and the soil matrix, ultimately governing preferential flow in structured soils. However, quantitative experiments of the flow exchange dynamics in the soil around an earthworm burrow (drilosphere) are limited. Here, we developed an experimental setup to mimic preferential flow within a biopore while monitoring the water distribution by neutron radiography. The objective was to quantify the evolving water content distribution in the biopore with surrounding drilosphere samples from Bt and C horizons. We hypothesize that the distribution of clay and organic material across biopore surfaces in Bt and C horizons impacts the biopore coating material formation (i.e., composition and thickness), thereby influencing its role in controlling mass exchange during preferential flow events. The experimental setup was proposed for in situ quantification of the preferential flow dynamics in an earthworm burrow (biopore) passing through Bt and C horizons of a Luvisol developed on glacial till. The neutron radiographies were obtained at the NEUtron Transmission Radiography facility of the Paul Scherrer Institut, Villingen. The volumetric water content was quantified by calibrating gray scale values from radiography images with defined water contents using stepped aluminum cuvettes attached to the samples; image analysis was used to determine sorptivity at the biopore–soil matrix interface. The water exchange flow was found to be smaller for coated biopore surface in the more clayey Bt horizon, as compared to the coarser textured carbonaceous subsoil C horizon, where higher sorptivity intensified mass exchange. The method most realistically captured the dynamics of the vertical preferential flow process in a biopore, including lateral exchange flow, by determining the water sorptivity of the drilosphere. In contrast to mini‐infiltrometer data, biopore–matrix exchange flow was here more restricted for Bt in contrast to carbonaceous C horizon. Since both horizons showed the presence of coating material on their burrow walls, the results suggest further analysis of their contrasting properties and structures.

Improved Collision Probability Analysis of the Double-Heterogeneity Problem Based on Matrix Chord Length Correction

Dispersion fuel exhibits excellent safety performance and effectively reduces the risk of radioactive leakage, making it widely applied in high-temperature gas-cooled reactors (HTGRs) and other advanced nuclear reactors. The presence of stochastic media in dispersion fuel leads to the challenging double-heterogeneity problem in neutron transport calculations. Hébert proposed a collision probability analysis model for treating stochastic media, which has been implemented in the DRAGON5 code. As one important basis of derivation, it is assumed in the Hébert model that the neutron transmission probability is identical to the neutron escaping probability in matrix material. In this paper, it is figured out that the assumption is not rigorous for realistic stochastic media. Then, an improved approach based on the Hébert model is proposed to take into account the realistic chord length distribution as well as to ensure the conservation and reciprocity of collision probabilities. The proposed methodology has been implemented in the HTGR lattice physics code XPZ. By numerical analysis against Monte Carlo reference solutions, it is demonstrated that the improved Hébert model with chord length correction gives good accuracy for addressing realistic double-heterogeneity problems.

Neutron transmission analysis in borated polyethylene, boron carbide, and polyethylene: Insights from MCNP6 simulations

This study delves into the neutron transmission properties of three pivotal materials: borated polyethylene, boron carbide (B4C), and polyethylene, employing the advanced MCNP6 (version 6.2) Monte Carlo code. The primary objective was to discern the impact of boron content and material density on neutron transmission efficacy. Results highlighted the formidable neutron shielding prowess of B4C, which exhibited the lowest neutron transmission factor among the studied materials. Borated polyethylene samples revealed an inverse correlation between boron concentration and neutron transmission, emphasizing boron's paramountcy in neutron attenuation. In contrast, polyethylene showcased notable neutron moderating capabilities, necessitating synergistic integration with materials possessing higher neutron absorption cross-sections for superior neutron shielding. Material density and the presence of other elemental constituents further influenced neutron transmission. The research provides a comprehensive understanding of neutron-material interactions, setting the groundwork for enhanced safety measures in nuclear applications.

Neutron Transmission CB-KID Imager Using Samples Placed at Room Temperature

Neutron Transmission CB-KID Imager Using Samples Placed at Room Temperature

Optimization of polycapillary lens for neutron focusing

Using various neutron focusing techniques to obtain the focal spot with small size and high gain at the sample location is highly desired in micro-area analysis. Our previous work fully demonstrated the feasibility of polycapillary lenses in the field of neutron focusing; however, the focusing performance of the lens fell short of expectations. This study attempts to optimize the lens by adjusting the design parameters of the lens in order to achieve better focusing performance and higher gain. In addition, microscopic observations revealed that the previously developed lens has some defects in its microstructure, which are not conducive to neutron transmission. In this paper, a manual sorting process was added to the lens manufacturing process to obtain a lens with improved microstructure. The focusing performance of the optimized lens was tested using quasi-parallel neutron beams from China Spallation Neutron Source (CSNS). The experimental results confirm the effectiveness of the optimization strategies, which has a significant guiding property on the development of lenses.

Evaluation of shielding and interaction properties of different stainless steel alloys for nuclear power plant shielding

This study evaluated the gamma and neutron-shielding parameters, electron and charge particles’ interaction properties of three different stainless steel-based alloys (2507SS, 304SS and AISI1018) with densities of 7.82, 7.81 and 7.82, respectively for deployment in nuclear reactor construction. Gamma shielding parameters were evaluated with Phy-X/PSD, Py-MLBUF and GRASP computer programs within the energy range of 0.2 MeV to 15 MeV. The relative deviations (RD %) between two of these computational platforms were calculated. Fast Neutrons Effective Removal Cross-Sections (FNRCS) and Macroscopic Removal cross-section (MRCS) values for fast neutrons were calculated with Phy-X/PSD and MRCS software, while thermal/fast neutron attenuation factors were calculated with NGCals software. The continuous-slowing-down approximation (CSDA) range and total stopping power (TSP) values of H+ and He++ ions were estimated with SRIM Monte Carlo code in a wide energy range of 1–20 MeV. The projectile range of electrons was computed with ESTAR NIST software within the kinetic energy of 0.01–1000 MeV. Gamma-ray, thermal neutron and fast neutron transmission factors (TFs) values were calculated for all samples for a range of well-known energies. From the obtained results, all evaluated parameters were dependent on the composition of shielding materials, the type of radiation and the photon energy of the radiation. The maximum MAC values obtained from the three alloy samples are 0.1505 cm2/g for 2507SS, 0.1453 cm2/g for 304SS and 0.1459 cm2/g for AISI1018 at 0.02 MeV. These results projected 2507SS as a better shielding alloy when compared with other investigated alloy samples. Considering the closeness of densities of investigated alloy samples the shielding and interaction properties of all alloy samples gave excellent and similar results implying that all investigated alloys will effectively serve as a radiation-shielding material in a nuclear reactor.