Thermal Neutron Cross Section Research Articles

Physical, chemical and radiation shielding properties of metakaolin-based geopolymers containing borosilicate waste glass

Upcycling waste borosilicate glass (BS) is a viable and sustainable way of preserving the natural resources required for the production of new materials and ensuring environmental sanitation and safety. This study was aimed at investigating how the addition of BS influences the physical, chemical, and mechanical properties and shielding performance against neutrons and charged particles on metakaolin-based geopolymer G. Four batches of G infused with 10%, 20%, and 30% wt BS samples and coded as G, G-10BS, G-20BS, and G-30BS, respectively, were fabricated based on the solid-state reaction method. The samples were characterized accordingly using experimental and theoretical procedures. X-ray diffraction analysis of the materials showed that the addition of BS suppressed crystal peaks in the geopolymers. The Vickers hardness values were calculated as 554, 503, 517, and 528 HV under a 0.1 kg load. The values of fast neutron macroscopic cross-section increased from 0.0601 cm−1 to 0.0706 cm−1 as BS content increased from 0 to 30 wt%. The coherent, incoherent, absorption, and total interaction probabilities of thermal neutrons increased in the geopolymer samples. For electrons, the stopping powers were in the range of 1.575–13.17 cm2/g, 1.571–13.15 cm2/g, 1.571–13.16 cm2/g, and 1.571–13.18 cm2/g for G, G-10BS, G-20BS, and G-30BS, respectively. The projected ranges of carbon ions in G, G-10BS, G-20BS, and G-30BS are within the limits 0.069–17.87 μm, 0.064–16.99 μm, 0.064–16.99 μm, and 0.057–15.24 μm. The trend of the projected ranges of electrons, protons, α-particles, and C ions with kinetic energies within 15 keV and 15 MeV was mostly in the sequence G > G-10BS > G-20BS > G-30BS. This order was consistent with the density and BS content of the geopolymer samples. The addition of BS to the metakaolin-based geopolymer improved the fast neutron moderating capacity and thermal neutron cross-sections and reduced the range of charged particle radiation in the geopolymers. The neutron shielding ability of G-30BS was found to be better than that of some existing shielding materials. The BS-infused concrete can be useful for the production of concrete used for radiation control.

Zircoborate glass-ceramic composite reinforced with aluminum oxynitride: Microstructural evolution, physical properties, and charged radiation attenuation analysis

This study used the powder metallurgy method to create aluminum oxynitride (AlON)-reinforced zircoborate glass-ceramic composites. The solid-phase reaction method was used to synthesize AlON powder, which was then added to Zr-based glass-ceramic in different amounts (0, 4, and 8 % by weight) to make it stronger and harder. The prepared materials were characterized for their structural, physical, and radiation shielding properties. The density of GZr8 (AlON-free) had a value of 2.9 g/cm3, and the sintered glass-ceramic density with 4 % wt. AlON reinforcement (GZr8Al4) reached 3.05 g/cm3. The highest density of 3.11 g/cm3 was obtained for the GZr8Al8 sample with 8 % AlON wt. The Vickers hardness for GZr8, GZr8Al4, and GZr8Al8 was measured to be 4.89, 5.94, and 6.33 GPa, respectively. The AlON-reinforced GZr8 samples had fast neutron (FN) removal cross-sections of 0.0796 cm−1 and 0.0824 cm−1 for AlON content of 4 and 8 wt%, respectively. Generally, the coherent scattering cross-section (σcoh), incoherent scattering cross-section (σinc), and absorption cross-section (σabs) of thermal neutrons (TNs) increase from 0.047856 cm−1, 0.02072 cm−1, and 0.01498 cm−1, respectively for GZr8 to 0.53124 cm−1, 0.02074 cm−1, and 0.01795 cm−1, for GZr8Al8. AlON was proven to improve the ability of the Zr-based glass-ceramics to attenuate FNs and TNs. Stopping powers of electrons, protons, alpha particles and carbon ions increase with the AlON content of the materials. Comparatively, the range of charged radiation in AlON-rich samples is lower. The inclusion of AlON in the matrix of the Zr-based glass-ceramics makes it more efficient at absorbing charged radiation and can be used in ion beam or hadron therapy, ion beam analysis, or other charged radiation environments as an absorber.

Evaluation of shielding performance and light emission characteristics of luminescent radiation shields using Gd2O2S

ABSTRACT Gadolinium oxysulfide (Gd2O2S: GOS) has a large absorption cross-section for thermal neutrons and a mass density similar to that of iron used for shielding against X and γ-ray radiation. During GOS scintillator production process, its waste is also generated. If GOS waste can be used as an effective shielding material, it is worthwhile as beneficial use of the waste. We developed a novel radiation-shielding material using GOS waste. The developed shielding material can shield X, γ-rays and thermal neutrons more efficiently than conventional shielding materials and emit visible light due to leaked radiation when the shielding material is damaged, which subsequently be used to check the integrity of shielding using visible light. In this study, two types of shields, concrete and mortar with GOS as the aggregate and epoxy resin with GOS were prototyped and tested for radiation transmission. Results indicate that GOS shielding has a 10% higher shielding ability for X and γ-rays than heavy concrete with iron as aggregate. Additionally, the same shielding effect for thermal neutrons was attained at 1/10 the thickness of heavy concrete. Thus, using GOS waste is not only cost effective but also contributes to a circular economy.

Recycled cathode ray tube waste glasses for radiation shielding applications: Role of Na2CO3

The aim of the study is to evaluate the ability of cathode-ray tube funnel waste glasses (CRT-F) to shield photons (gamma and X-ray) and other forms of radiation with rest masses (i.e., neutrons and charged particles) after their Pb content was reclaimed. The lead (Pb) content of cathode-ray tube funnel waste glasses (CRT-F) was reduced using Na2CO3 as the reducing agent CRT-F(Na2CO3). CRT-F and CRT-F(Na2CO3) were produced using the melt and quench processes using the powder of pristine funnel glass from discarded cathode ray tube glass and the reduced glass powder, respectively, as the starting materials. The X-ray fluorescence technique was adopted for determining the chemical composition of CRT-F and CRT-F(Na2CO3) glasses. The parameters relating to the gamma photon, fissile neutron, moderated neutron, slow neutron, and charged radiation (β, H+, He2+, and C6+) attenuation abilities of pristine CRT-F and CRT-F(Na2CO3) were evaluated. The FLUKA simulation code and XCOM were used to estimate the mass attenuation coefficient of the glasses. The stopping power (Sp) and range (R) of β, H+, and He2+ were evaluated by using the NIST data-based calculators (ESTAR for β, PSTAR for H+, and ASTAR for He2+) while Sp and R data for C6+ was determined using SRIM. The cross section for fast, thermal (25 meV), and slow neutrons (0.1–10 meV) was also estimated. The Pb reclamation from the CRT-F drastically increased the half value layer of photons by at least a factor of 4 at 0.1 MeV and by more than 90 % at 10 MeV. The CRT-F interacts more with fissile and thermal neutrons compared to CRT-F(Na2CO3). Although, CRT-F showed a better ability to attenuate all the radiation types considered, both glasses are better absorbers of radiation (especially photons) than some common shields. The reduction of Pb weight in CRT-F resulted in a drastic reduction in the ability of the glass to attenuate radiation; the resulting glass (CRT-F(Na2CO3)) is however, an environmentally safer glass shield.

Obtaining Boron Carbide and Nitride Matrix Nanocomposites for Neutron-Shielding and Therapy Applications

The very high capture cross-section of (epi)thermal neutrons by the boron isotope 10B makes elemental boron and its compounds and composites prospective for serving as materials intensively interacting with neutron irradiation. In their nanostructured form, boron-rich materials reveal properties that improve their radiation-performance characteristics. In this regard, new technologies have been proposed for the synthesis of nanocomposites with matrices of boron carbide B4C and hexagonal boron nitride h-BN. For the first time, boron carbide-tungsten and hexagonal boron nitride–(iron,magnetite) composites were obtained, respectively, in the form of layered/sandwich structures of components B4C and W and h-BN nanopowders coated/intercalated with magnetic nanoclusters of iron Fe or magnetite Fe3O4. Studying of their chemical/phase composition, structure/morphology, and some other properties leads to the conclusion that the developed B4C–W and h-BN–(Fe,Fe3O4) composites would be useful for solving important problems of boron-based neutron shielding and BNCT (Boron Neutron Capture Therapy), such as attenuating the gamma-radiation accompanying the absorption of neutrons by 10B nuclei and targeted delivery of 10B nuclei, as BNCT therapeutic agents, to tumor tissues using control by an external magnetic field, respectively.

Synthesis of Bi2O3 doping powder from CRT-screen waste glass: Physical, structural, and radiation attenuation properties

The physical, structural, mechanical, and radiation (gamma photon and neutron) shielding properties of CRT screen glass and Bi2O3 mixtures are presented in this study with the aim of adopting them as cheap, transparent, and environmentally safe glass shields for radiation protection purposes. Bi2O3-doped CRT screen glass samples are prepared by the ball milling and cold press methods and coded as TVS, TVS-Bi10, and TVS-Bi20, representing glass-ceramics containing 0, 10, and 20 wt% of Bi2O3, respectively. Vickers hardness values are 493, 534, and 564 HV for TVS, TVS-Bi10, and TVS-Bi20, respectively. Also, the values of densities are measured and recorded as 2.61, 2.78, and 3.06 g/cm3 for the TVS, TVS-Bi10, and TVS-Bi20. The values of the mass attenuation coefficients computed using FLUKA simulations and XCOM agree within an error margin of 1%. The values of the estimated photon shielding parameters show that shielding efficiency follows the order TVS < TVS-Bi10 < TVS-Bi20. In addition, the introduction of Bi2O3 has a positive impact to increase the ability of TVS to moderate fast neutrons The total cross-section for thermal neutrons in TVS, TVS-Bi10, and TVS-Bi20 is 1.1294, 1.1198, and 1.1932 cm-1, respectively. The best photon, fast, and thermal neutron absorber is TVS-Bi20 among the TVS-BiX glass-ceramics. The TVS-Bi20 sample can function better than some commercially available glass shields, light and heavy concrete, and other categories of composite materials such as glasses, rocks, and polymers. The TVS-Bi20 is thus recommended as an eco-friendly and efficient radiation shielding material.

Lead-free bismuth glass system towards eco-friendly radiation shielding applications

In this study, the gamma-ray photon, electron, proton, helium ion, carbon ion, thermal neutron, and fast neutron interaction parameters of the (70-x)(Bi2O3 + ZnO) + x(V2O5 + TeO2) + 20B2O3 + 6.5SiO2 + 2.5Al2O3 + 1(CeO2 + Sb2O3) glass system were estimated with the aim of assessing their radiation shielding potentials. The mass attenuation coefficients (MACs) of the glasses were estimated for photon energies between 0.015 and 15 MeV using the PHITS simulation code and XCOM software. The stopping powers of charged radiation such as electrons, protons, α-particles, and carbon ions were also estimated by Monte Carlo simulation of particle transport using PHITS. The narrow beam transmission simulated using PHITS yielded values of MAC that were not so different from those from XCOM. The MACs of B/V-ZBVSA1 vary from 0.039–90.747 cm2/g, while those of B/V-ZBVSA2, B/V-ZBVSA3, and B/V-ZBVSA4 vary from 0.035–69.724, 0.036–73.088, and 0.037–70.597 cm2/g, respectively. The HVL of the glasses varied from 0.011–2.56 cm, 0.019–3.69 cm, 0.017–3.31 cm, and 0.017–3.36 cm for B/V-ZBVSA1 – B/V-ZBVSA4, respectively The effective atomic number varied within the limits: 28.36–76.06, 19.67–67.91, 22.14–68.70, and 22.39–68.08 for BV-ZBVSA1 – BV-ZBVSA4, respectively. The stopping powers (Sp) of electrons in BV-ZBVSA1, BV-ZBVSA2, BV-ZBVSA3, and BV-ZBVSA4 were in the range 1.219–8.043 MeVcm2/g, 1.295–9.306 MeVcm2/g, 1.264–8.859 MeVcm2/g, and 1.257–8.793 MeVcm2/g correspondingly. Comparatively, the Sp of protons, α-radiation, and carbon ions follows the order: BV-ZBVSA2 > BV-ZBVSA3 > BV-ZBVSA4 > BV-ZBVSA1. Also, the value of ΣR for BV-ZBVSA1, BV-ZBVSA2, BV-ZBVSA3, and BV-ZBVSA4 was 0.1171, 0.1126, 0.1133, and 0.1103 cm−1 accordingly, while the corresponding total cross-section of thermal neutrons was 67.6227, 46.8356, 59.1179, and 56.6994 cm−1. The present glasses showed superiority in absorbing photons and neutrons compared to some existing shields and previously recommended glasses. The present glassy samples, thus, can be adopted for radiation protection purposes, among others in radiation facilities.

Optical properties and radiation shielding performance of tellurite glassy composites: Role of rare earth oxides

In this research, the synthesis, optical parameters, and radiation absorption quantities of 50TeO2–15Na2O—(20-x)PbO—15ZnO—xEu2O3 glasses were reported. The glasses were prepared by the well-known melting and quenching method using high purity chemicals. The glasses were coded as TPEu-0, TPEu-1, TPEu-2, and TPEu-3 for glasses containing 0, 1, 2, and 3 mol% of Eu2O3, respectively. The density of the glasses was determined by the Archimedes method. The optical parameters of the glasses, such as the refractive index n, molar refractivity Rm, molar polarizability αm, reflection loss RL, optical transmission T, metallization criterion M, static dielectric constant εstatic, and optical dielectric constant εoptical were estimated by theoretical methods. The stopping powers and range of electrons, protons, α-particles, and carbon ions in the TPEu-X glasses for particle kinetic energies between 15 keV and 15 MeV were computed using the ESTAR, PSTAR, ASTAR, and STRIM software. In addition, the fast neutron (FN) removal cross section (FNRCS), thermal neutron (TN), and total cross-section (σtot) for the prepared glasses were evaluated theoretically using established models. The estimated optical parameters showed strong relationship with the chemical content of the glasses. Glass density and molar volume increased as Eu2O3 content increased. The Rm, αm, RL, εstatic, and εoptical vary from 16.24 to 17.20 cm3/mol, 6.44 × 10−24 cm3‒17.20 × 10−24 cm3, 0.17–0.19, 5.71–6.55, and 4.71–5.55 respectively, with TPEu-0 and TPEu-1 having the maximum and minimum values, respectively. Generally, the increase in Eu2O3 concentration increased the charged radiation shielding ability of the glasses. The increment in the Eu2O3 content of the glasses brought about an improvement in the thermal and fast neutron absorption probabilities of the glasses.

Radiation shielding and nuclear security functions of Na2O–Fe2O3–B2O3–La2O3 glass system

The radiation shielding parameters of Fe3+ hybrid 20Na2O–5.0Fe2O3–(75- x) B2O3–xLa2O3 (where x = 0.0 (NBFL1), 0.5 (NBFL2), 1.0 (NBFL3), 1.5 (NBFL4), and 2.0 (NBFL5) mol%) were estimated and analysed in this report. The photon, neutron, and charged radiation interaction quantities of the glasses were estimated with the aid of the FLUKA Monte Carlo code, the XCOM calculator, and well-known theoretical expressions. The photon and charged particle shielding quantities were estimated for the particle's energy up to 15 MeV. The mass attenuation coefficient ranged from 0.0189 to 6.1457 cm2/g for NBFL1, 0.0194–7.3041 cm2/g for NBFL2, 0.0199–8.4231 cm2/g for NBFL3, 0.0204–9.5046 cm2/g for NBFL4, and 0.0209–10.5504 cm2/g for NBFL5. At 15 MeV the half value layer was 18.08 cm, 16.54 cm, 14.63 cm, 14.05 cm and 13.03 cm for NBFL1, NBFL2, NBFL3, NBFL4 and NBFL5, respectively. NBFL5 has the highest shielding efficiency for charged radiation due to its higher density and La2O3 content. The value of ΣR for NBFL1, NBFL2, NBFL3, NBFL4, and NBFL5 is 0.0842 cm−1, 0.0884 cm−1, 0,0961 cm−1, 0.0965 cm−1, and 0.1004 cm−1, respectively. There was also a steady increase in the total cross section for thermal neutrons as the density and La2O3 content of the glasses increased. In addition to advanced optical functions, the NBFLX glasses possess good radiation absorption qualities, which makes them attractive for radiation absorption roles in radiation environments, among other functions.

Effect of Deposition Temperature on the High‐Temperature Corrosion Resistance of CrMoNbZr Medium‐Entropy Alloys Coatings

A near‐equimolar CrMoNbZr coating with low thermal neutron cross sections is prepared on zircaloy‐4 substrate by radio frequency magnetron sputtering. The microstructure, mechanical properties, and surface wettability of the coatings prepared at different deposition temperatures are studied. The high‐temperature corrosion resistance of the CrMoNbZr coatings is investigated by a 40 d autoclave corrosion test with pure water containing 3.5 ppm lithium, 1000 ppm boron at 320 °C. The results indicate that the CrMoNbZr coatings are mainly composed of body‐centered cube‐structured nanocrystals, and the coating deposited at 300 °C has superior high‐temperature corrosion resistance and mechanical properties. Finally, the mechanism of high‐temperature corrosion is elucidated.

Design, synthesis, and characterization of Pb-free tellurite glasses for radiation shielding applications

The melt and quench technique was used to create a new TeO2–Li2O–MoO3 glass system with the objective of providing alternative and new shielding materials for radiation safety purposes. The synthesized glasses were studied by means of their physical, thermal, and radiation shielding characteristics. It was demonstrated using Raman and FTIR spectroscopy that the insertion of Li2O and MoO3 resulted in a superimposed reduction of all Te units, including TeO3+1 and TeO3 with non-bridging oxygen, via the formation of Te(short)-O-Mo and a coordination change from MoO4 to MoO6. The densities of the prepared glasses were relatively high, having values between 4.48 and 5.11 g/cm3. Compared with TeO2-pure glass, the values for Tg and Tx substantially decrease with Li2O and MoO3 insertion. However, among prepared samples, there is only a small fluctuation in both characteristic temperatures, suggesting high thermal/composition stability. FLUKA simulations were used to estimate the photon, proton, electron, α-particle, and carbon ion shielding parameters for beam energies ranging from 15 to 15,000 keV. With changes in the chemical designation of the synthesized glasses, the mass attenuation coefficient of photons and stopping powers of charged radiation changed, gradually increasing with glass density and decreasing with Li2O and MoO3 insertion. In addition, the ability of the glasses to moderate fast neutrons declined with increases in the MoO3 and Li2O weight contents. On the contrary, the total cross section of thermal neutrons improved as glass density decreased and Li2O and MoO3 content increased. The synthesized glasses, through the evaluated parameters, showed better radiation shielding capacity compared to existing shields such as concrete and commercial RS-series glass shields.

Hot hydrogen testing of direct current sintered NbC and (ZrNbW)C for nuclear thermal propulsion

Refractory carbides are promising materials for use in nuclear thermal propulsion, either as the primary material in nuclear fuel elements or as protective coatings. In particular, NbC and ZrC have high melting points, relatively low vapor pressures at elevated temperatures, and low thermal neutron cross sections. In this study, NbC and (Zr0.4Nb0.4W0.2)C were fully consolidated by direct current sintering, and relative densities of 98% were achieved. The samples were characterized by scanning electron microscopy and X-ray diffraction before and after hydrogen testing. The samples were exposed to hot flowing hydrogen gas at a pressure of 1 atm and a flow rate of 8 SLPM, at temperatures up to 2500 K, and for durations up to 3 h. The mass loss rate (MLR) of NbC decreased with increasing exposure, before reaching a steady state value of about 1 mg/m2s. For (ZrNbW)C, a comparison was made by exposing samples to either hot flowing hydrogen or argon gas. The mass loss rates decreased with increasing exposure to either gas, but the final steady-state MLR was one to two orders of magnitude greater for those samples exposed to hydrogen. As evidenced by X-ray data, small amounts of residual oxygen in the (ZrNbW)C samples were preferentially removed upon hydrogen exposure.

Evaluation of Thermal Neutron Scattering Law and Cross Sections for Calcium Hydride

Presented here are the calculated thermal scattering law (TSL) and thermal neutron scattering cross sections for Calcium Hydride, hereafter referred to by its chemical symbol CaH2. The only other such data prior to this evaluation are thermal neutron scattering libraries in the JEFF database, which do not fully capture the scattering physics of the CaH2 system. The data in this evaluation are calculated from first principles; Density Functional Theory (DFT) is used to calculate the phonon density of states (DOS), which is the primary input required to derive the TSL. The TSL and cross sections have been evaluated for the three non-equivalent atom cites in the CaH2: Ca, H1, and H2. Each evaluation has been submitted to the NNDC for consideration in the next ENDF/B database release.

Neutron thermal cross sections of 3D-printing organic polymers using the Average Functional Group Approximation

We provide a worked example on how to obtain the total neutron scattering cross section of organic polymers at thermal neutron energies by means of the Average Functional Group Approximation. Within this framework, a polymer is rationalised as composed of hydrogen atoms belonging to specific functional groups,therefore taking into account the material-specific incoherent inelastic scattering contributions to the thermal cross section. Such simplified model allows the calculation of thermal neutron cross sections for a broad variety of materials of interest in neutron applications, e.g., sample containers and shielding components. We discuss in detail the case of ABS, often available as a 3D-printing material, whereby three polymers (acrylonitrile, butadiene, and styrene) can be combined in several mass percentages depending on the final application. Results are obtained using the NCrystal module already featuring the Average Functional Group Approximation.

Europium doped Gd2O3 and GdBO3 scintillators for thermal neutron detection

Neutron detectors used in various applications in nuclear security and nuclear safety are mostly based on the 3He technology. Unfortunately, in the last few years, the market of 3He has encountered huge problems in matching the supply and demand leading to an exponential increase in the price and a serious strategic problem of resources. To guarantee the availability of detection systems for nuclear security, the last decade has been driven by the quest for exploring alternative technologies to replace 3He based detectors. Gadolinium (natGd) is a promising rare earth element which has the largest capture cross-section for thermal neutrons among all stable elements due to the contributions of the isotopes 155Gd and 157Gd. This paper describes the fabrication of Gd2O3:Eu3+ and GdBO3:Eu3+ phosphors as scintillators for thermal neutron detection. The samples were evaluated using photoluminescence, SEM analysis, and pulse height spectra recorded from a D–D neutron source. The recorded spectrum was compared to a FLUKA simulation of the characteristic K X-ray emission following neutron capture.

Nuclear energy system modelling application package: functional overview and examples

The paper provides a brief description of the functionality of the nuclear energy system modelling application package (NESAPP), including a list of modelling objects, key assumptions and areas of application. NESAPP consists of the following main modules: NUDAPS (a module for calculating thermal neutron cross-sections, resonance integrals and one-group neutron cross-sections, and associated uncertainties), NUCLEX (a module for calculating the evolution of the nuclide composition and characteristics of nuclear fuel in reactors and at the nuclear fuel cycle front-end and back-end steps), NUCAB (a module for adjusting isotopic composition and blending), FANES (a module for analysing material flows and integrating data in nuclear energy evolution scenarios), ECNES (a module for assessing economic performance metrics for the nuclear energy evolution scenarios). Each of the modules is a calculation tool that can be used as independent or integrated into the software for technical and economic modelling of nuclear energy systems. Various calculation models are implemented in the modules, allowing users to evaluate the methodological component of the calculation uncertainty in scenario modelling studies and the functionality for assessing the impact of initial data uncertainties on the resulting indicators. The authors also provide some examples of applying NESAPP.

Effect of MoO3 on Na2O–B2O3–CdO–ZnO glasses: Applications in optoelectronics, communication devices, and radiation shielding

The current study focuses on the physical and structural aspects of B2O3–CdO–ZnO–Na2O glasses that have been reinforced by MoO3 ions. The samples were experimentally prepared using a melt quenching procedure, and then the absence of strong peaks in the XRD spectra confirms their amorphous nature. A variety of optical, Raman, and radiationshielding qualities were evaluated and described. The structural changes in the material were investigated using FTIR and Raman spectra. The density of the samples rose when MoO3 was added, implying that BO3 was transformed into BO4 units. The transition of Mo6+ to Mo5+ ions is responsible for the reduction in band gap in these glasses. Because of the vacancy generated by the oxygen units, the Urbach energy increased, implying the generation of additional defects. The presence of diverse structures present in the structure of the glass was revealed by FTIR and Raman spectra, such as MoO6, BO3, BO4, metaborates, pyroborates, and so on. The transition of BO3 to BO4 units, which is crucial in density analyses, was also corroborated by Raman spectra. Using FLUKA simulation and XCOM estimations, the mass attenuation coefficient of the glasses improved with the amount of MoO3. At 15 keV, the half-value layer of 0.0103 cm, 0.01019 cm, 0.01016 cm, 0.0101 cm, and 0.01005 cm was obtained for glasses with an MoO3 concentration of 0, 0.5, 1, 1.5, and 2.0 mol% respectively. The optimum MoO3 concentration for fast and thermal neutron cross-sections was 0.5 and 2.0 mol% respectively. The glasses are viable materials for environmentally friendly and transparent radiation barriers in nuclear facilities.

Radiation shielding capacity of Li2O-SiO2/GeO2 glasses doped with rare earth oxides: Nuclear security applications

In the present work, the ionizing radiation (namely, gamma-rays, charged particles, and neutrons) attenuation parameters of Li2O–SiO2 (denoted as LS) and Li2O–SiO2–GeO2 glasses doped with Y2O3 and In2O3 (LGY and LGI) glasses were estimated and analyzed with the aim of exploring their shielding potentials. The gamma-ray transmission attenuation coefficient was simulated by the FLUKA code and also estimated by XCOM. Also, the stopping powers and ranges of electrons, protons, α-particles, and carbon ions were estimated for particle kinetic energies within 15–15000 keV. The fast and thermal (25 meV) neutron cross-sections were estimated for the three glasses. At 15 MeV, the mass attenuation coefficient is 0.019, 0.021 and 0.022 cm2/g while the linear attenuation coefficient magnitude varied within the range of 0.047–11.564 cm−1, 0.061–32.008 cm−1 and 0.068–38.122 cm−1 for LS, LGY, and LGI, respectively. The effective atomic number varied from 8.21 to 12.00 for LS, 8.71–22.71 for LGY, and 8.78–26.37 for LGI. Likewise, the charged particle shielding ability of the glasses was in the order: LS < LGY < LGI. This trend was conserved for the neutrons’ cross-sections. Compared to conventional and recently researched shields, LS, LGY, and LGI present viable alternative glass shields in nuclear radiation applications.

Total thermal neutron cross section measurements of yttrium hydride from 0.0005 - 3 eV

Yttrium hydride serves as a neutron moderator material that enables compact, high temperature nuclear reactors. However, in order to accurately design and simulate a nuclear system relying upon yttrium hydride, the fundamental nuclear data of yttrium hydride must be well understood. Thermal neutron scattering law (TSL) evaluations represent an important aspect of nuclear data as thermal scattering can drastically alter the neutron multiplication factor of a system. Therefore, to support evaluation and validation of thermal neutron scattering for yttrium hydride, researchers at Rensselaer Polytechnic Institute (RPI) performed total thermal neutron cross section measurements for YH1.68 and YH1.85 over the energy range of 0.0005 - 3 eV. These measurements represent the first total cross section measurements for yttrium hydride that encompass the entire thermal region. Comparisons were made against the ENDF-B/VIII.0, Zerkle & Holmes and Oak Ridge National Laboratory TSL evaluations, where generally good agreement was found.

Effect of the B4C content on microstructure, microhardness, corrosion, and neutron shielding properties of Al–B4C composites

Aluminum–Boron Carbide (Al–B4C) metal matrix composites (MMCs) are frequently studied because of their high thermal conductivity and ability to absorb neutrons, as an alternative material for the manufacture of baskets in spent fuel casks in the nuclear industry. Keeping in mind, this study aimed to produce Al–B4C MMCs and examine their properties to shield against neutron radiation. The Al-xB4C (x = 5, 10, 15, 20, 25, 30, 40, 45, and 50 wt%) composite powders were prepared by a high-energy planetary ball milling. The prepared powders were compacted with uniaxial cold compaction at 500 MPa and sintered in a tube furnace at 600 °C for 1 h under an Argon atmosphere. The microstructure of the obtained MMCs was characterized by using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). The XRD result indicates that the synthesized Al–B4C MMCs have a predominant phase of Al and B4C. The SEM images show the distribution of B4C reinforcement inside the Al matrix. The relative density, measured by the Archimedes principle decreased from 99.95 to 91.55% as the B4C ratio increased. The microhardness value was increased in the beginning from 44 to 75.5 HV and then decreased to 36.5 HV as the B4C ratio increased. The polarization curves show that the corrosion current density and the corrosion rate were significantly increased from 52.25 to 375.36 μA cm−2 and 0.573–4.121 mm/year as the B4C ratio increased. The simulation results, computed by the MCNP6.2 program show that the thermal neutron cross-section and the fast neutron cross-section was increased from 4.5 to 40.1 cm−1 and 0.13 to 0.15 cm−1 as the B4C ratio increased. Further, the neutron equivalent dose rate was calculated experimentally with the 241Am–Be neutron source and the data was completely agreed with the computational results.