Neutron Cross Sections Research Articles

Temperature distribution during welding measured by neutron imaging

This study was carried out to investigate the neutron transmission signal as a function of sample temperature during a welding process. A theoretical description that includes the Debye-Waller factor was used to describe the temperature influence on the neutron cross-sections. Neutron imaging using a monochromatic beam helps to observe transmission variations related to the material temperature. In-situ neutron imaging of welding experiments show the distribution of the temperature in bulk steel samples. The performed finite element modelling of expected temperature distributions shows good agreement with the obtained experimental data.

The interacted characteristically study of alpha, proton, neutron, and gamma radiation against Gd2O3-Na2O-Al2O3-P2O5 glass system doped with Dy3+: Theoretical investigation

The radiation shielding properties and fast neutron removal cross-sections of Na2O-Gd2O3-Al2O3-P2O5 glass system doped with Dy3+ were calculated by theoretical approach using the Phy-X program. However, charged particles were simulated via SRIM coding. The result found that density and molar volume of the samples increase with increasing of Dy2O3 concentrations. For radiation shielding properties, increasing Dy2O3 concentrations impact with the mass attenuation coefficient (MAC), effective atomic number (Zeff), and effective electron density (Neff) and also corresponds to the gamma energy ranges. In other hands, increasing density and the MAC are not result directly with neutron cross-sections. For charged particles, the total mass stopping power (TMSP) of both particle as proton and alpha particles depend on the medium that charged particles traversed and also rely on number of ions. HIGHLIGHTS Calculate and analyst the rediation shielding parameters by Phy-X program. The charged particles were simulated via SRIM coding. The study of radiation shielding properties Na2O-Gd2O3-Al2O3-P2O5 GRAPHICAL ABSTRACT

Verification of T Invariance in the Total Interaction Cross Section of Neutrons with Unpolarized Nuclei Using the Polarization–Asymmetry Theorem

Verification of T Invariance in the Total Interaction Cross Section of Neutrons with Unpolarized Nuclei Using the Polarization–Asymmetry Theorem

Theoretical Investigation of the Influence of Different Heavy Metal Oxides Modifiers on ZnO-Bi2O3-B2O3-SiO2’s Photon- and Neutron-Shielding Capabilities Using the Monte Carlo Method

Radiation has become an essential part in medicine and researchers are constituently investigating radiation shielding materials that are suitable for different medical applications. Glass, due to its properties, has been considered an excellent radiation shield for such applications. One of the most common glasses used as a radiation shield is the ZnO-Bi2O3-B2O3-SiO2 anti-radiation glass. Heavy metal oxides have many desirable properties such as high density, transparency to visible light, stability in air and water, high interaction cross section, high infrared transparency, and good absorption of radiation, which make them desirable to be used as modifiers with anti-radiation glass. Research has been focusing on environmentally friendly shielding material which leads to non-lead modifiers such as Na2O, Al2O3, MgO, TiO2, SrO, Sb2O3, and BaO, which have become more desired than PbO. So far, ZnO-Bi2O3-B2O3-SiO2’s photon shielding properties have been studied experimentally with the addition of BaO at certain energies only. In this work, different heavy metal oxides are added as modifiers to ZnO-Bi2O3-B2O3-SiO2 glass in order to investigate theoretically their effects on the shielding properties of the glass at a wide range of photon and neutron energies. Simulation is cost- and time-effective when it comes to investigating different compositions of glass and different modifiers with different weight percentages at any energy range for any type of radiation. Simulation could be considered the first step in order to identify the best mixture with the best weight fractions prior to any experimental investigations of other desired properties based on the needed application. In this work, the photon- and neutron-shielding capabilities of the ZnO-Bi2O3-B2O3-SiO2 anti-radiation glass is investigated with different weight fractions of heavy metal oxides at wide photon and neutron energy ranges. Geant4, which is a Monte Carlo-based powerful toolkit, is used to find the mass attenuation coefficients (µm) of photons, as well as the effective removal cross sections (ΣR) of neutrons, of all the investigated samples in the studied energy range.

241Am Neutron Capture Cross Section Measurement using the NaI(Tl) spectrometer of the ANNRI beamline of J-PARC

ABSTRACT The neutron capture cross-section of 241Am was measured from 10 meV to about 500 keV using the NaI(Tl) spectrometer of the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) beamline in the Materials and Life Science Experimental Facility (MLF) of the Japan Proton Accelerator Research Complex (J-PARC). The total energy detection principle was applied in conjunction with the pulse-height weighting technique to derive the neutron capture yield. The present cross-section results were normalized using a 197Au sample measurement by applying the saturated resonance method. The thermal cross section was measured to be 708 ± 22 b, in agreement with the evaluation in JENDL-5 of 709 b within uncertainties. In the keV region, the evaluated libraries of JENDL-5, ENDF/B-VIII.0 and JEFF-3.3 are in good agreement with the present results over 10 keV but show values lower by 30% to 70% below that energy. Moreover, the results of a resonance shape analysis of the resolved resonance region are also provided in the present paper.

Elucidating of the influence of tin and cadmium in the Bi–Pb alloy on its thermal and nuclear properties as a coolant in fast neutron reactors

The purpose of this research is to investigate the physical properties and shielding characteristics of prepared (Bi–Pb) and (Bi–Pb–Sn–Cd) alloys, to be used as a coolant in a liquid metal nuclear fast reactor. The physical properties density, kinetics and mechanism of melting applying the reverse Johnson–Mehl–-Avrami rule and reduced time plot were applied, showing that melting processes take place in two dimensions, enthalpy and heat capacity were measured on the other hand for the prepared alloys within the solid and liquid states, respectively. Moreover, the results indicated that the bulk density and melting temperature of (Bi–Pb–Sn–Cd) alloys are lower than (Bi–Pb) alloys. Likewise, the thermal properties of (Bi–Pb–Sn–Cd) alloys seemed more suitable applicants than that of (Bi–Pb) alloys. Additionally, eight different gamma-ray energy lines in the range 0.12178–1.40792 MeV were used to determine the mass attenuation for the alloy samples. Also, gamma radiation attenuation characteristics, the half-value layer and effective atomic number were determined using Phy-x/PSD software. Furthermore, three types of neutron energies (slow neutrons, total slow neutrons, as well as neutrons of energies greater than 0.4 eV) were used to determine the macroscopic neutron cross-sections for the prepared alloys. The calculated values of mass attenuation coefficients exhibited a very good agreement with the experimental values of those parameters. The values of neutron macroscopic cross sections at fast neutrons for (Bi–Pb) and (Bi–Pb–Sn–Cd) alloys are small, this means that there is no consequence on fast neutrons used in a nuclear breeder reactor. Finally, we may reach the conclusion that Bi–Pb–Sn–Cd alloys are a possible candidate to be used as a coolant in fast nuclear reactors, as well as it is most suitable as an outer cooling and shielding blanket vessel as well.

Applying approximate Bayesian computation to reduce uncertainty in multigroup 235-U cross-sections using ICSBEP experimental data

The multiplication factor (keff) and its uncertainty are critical design parameters in nuclear reactors. The keff uncertainty must be considered for operation, safety, and economic reasons. Consequently, reducing uncertainty in keff has been of interest to the nuclear industry for as long as nuclear reactors were designed. Two methods of reducing this uncertainty are in use – Generalized Linear Least Squares (GLLS) and A General Monte Carlo-Bayes Procedure for Improved Predictions of Integral Functions of Nuclear Data (MOCABA). Both methods have the ability to reduce the keff uncertainty by reducing uncertainty in cross-sections through the assimilation of measured critical systems’ or nuclear reactors’ operational data. The algorithms require means of simulating the experimental data and that the calibrated parameters are described by specific statistical distributions. GLLS is limited to linear models and multivariate normal prior and posterior, while MOCABA can use any (non-linear) model but is also limited to multivariate normal prior and posterior. This work implements a universal and rigorous algorithm called Sequential Monte Carlo – Approximate Bayesian Computation (SMC-ABC) for the same application. It is found that SMC-ABC, GLLS, and likely MOCABA give essentially the same results for the same problems. The paper also presents a reliable method for validating the results based on so-called “synthetic experiments”. Synthetic experiments are computationally generated data used in place of experiments. The method allows for verifying whether there is a risk of overfitting the calibrated parameters (cross-sections in case of this work) during the data assimilation. This novel method shows the potential of using SMC-ABC and other Bayesian calibration methods to generate improved general-purpose neutron cross-section libraries.

Measurements of the Neutron Total and Capture Cross Sections and Derivation of the Resonance Parameters of 181Ta

The neutron total and capture cross sections of 181Ta were measured at the Accurate Neutron-Nucleus Research Measurement Instrument of the Materials and Life Science Experimental Facility in the Japan Proton Accelerator Research Complex to improve the accuracy of the current resonance parameters. The total cross section was determined from the transmission measurement in the energy range from 0.2 to 150 eV. The capture cross section was derived from the capture yield using the pulse height weighting technique in the energy range from thermal to 150 eV. The thermal neutron capture cross section was measured as 21.3 ± 0.3 ( stat ) ± 0.6 ( sys ) b. The obtained transmission and capture cross section were simultaneously fitted using the resonance analysis code REFIT, and the resonance parameters for resonances below 150 eV were evaluated. The present resonance parameters were compared to reported measurements.

Measurement of the ^{181} Ta(n,gamma ) cross sections up to stellar s-process temperatures at the CSNS Back-n

The neutron capture cross section of ^{181} Ta is relevant to s-process of nuclear astrophysics, extraterrestrial samples analysis in planetary geology and new generation nuclear energy system design. The ^{181}Ta(n,gamma ) cross section had been measured between 1 eV and 800 keV at the back-streaming white neutron facility (Back-n) of China spallation neutron source(CSNS) using the time-of-flight (TOF) technique and hbox {C}_{6},hbox {D}_{6} liquid scintillator detectors. The experimental results are compared with the data of several evaluated libraries and previous experiments in the resolved and unresolved resonance region. Resonance parameters are extracted using the R-Matrix code SAMMY in the 1–700 eV region. The astrophysical Maxwell average cross section(MACS) from kT = 5 to 100 keV is calculated over a sufficiently wide range of neutron energies. For the characteristic thermal energy of an astrophysical site, at kT = 30keV the MACS value of ^{181}Ta is 834 ± 75 mb, which shows an obvious discrepancy with the Karlsruhe Astrophysical Database of Nucleosynthesis in Stars (KADoNiS) recommended value 766 ± 15 mb. The new measurements strongly constrain the MACS of ^{181}Ta(n,gamma ) reaction in the stellar s-process temperatures.

Advances in Polymeric Neutron Shielding: The Role of Benzoxazine-h-BN Nanocomposites in Nuclear Protection.

Given their substantial neutron capture cross-section, extreme hardness, and high chemical and thermal stability, boron-based materials are widely used as building blocks to protect against highly ionizing radiations such as gamma rays and neutrons. Indeed, uncontrolled nuclear radiation exposure can be highly hazardous to radiation workers and the public. In this sense, this work presents an extensive study and experimental evaluation of the nuclear shielding features of hexagonal-boron nitride (h-BN) based nanocomposite, where bisphenol-A based polybenzoxazine (BA-PBz) was used as matrix. The neutron shielding studies were carried out at the nuclear research reactor of Algeria NUR. The surface treatment of h-BN nanoparticles was confirmed by FTIR and XPS techniques. The curing behavior and the degradation phenomena of the nanocomposites were evaluated by DSC-TGA analyses. The distribution of h-BN nanoparticles within the polymer matrix was assessed by TEM and SEM. The results showed that the developed boron nitride-based nanocomposite exhibits intriguing shielding performances and good thermal stability. The DSC-TGA tests exhibit high degradation temperature that reach 279°C. The highest performances were obtained at an h-BN concentration of 7 wt%, where the macroscopic cross was found to be (Σ = 3.844 cm-1) with a screening ratio of (S = 96.12%), equivalent to a mean free path (λ) of 0.138 cm.

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.

Al2O3/GdAlO3 reaction sintered composites produced from milled powders: Microstructure, neutron-capture cross section, and mechanical properties

The incorporation of high thermal neutron capture elements in ceramic materials is relevant for the design of traceable proppants for obtaining information in petrology, such as the location and height of fractures in unconventional gas and oil wells. Among these, GdAlO3 has attracted interest because of its good thermo-mechanical properties. In this work, four Al2O3/GdAlO3 reaction sintered composites were prepared using Gd2O3 and Al2O3 powders as raw materials and high-energy ball milling followed by sintering at 1600 °C. This combination of ball milling and reaction sintering is considered as a mechanism to produce large bulk ceramics with different Al2O3/GdAlO3 compositions. The composite formulations range from pure alumina to eutectic ceramic. The obtained materials were characterized using X-ray diffraction for identifying the crystalline phases, Archimedes method for the textural properties, scanning electron microscopy for the microstructure, and Vickers indenter and three-point bending tests for the mechanical properties. In addition, the macroscopic capture cross-section Σ for thermal neutrons was estimated considering each composite formulation. It was found that the processing route is very suitable for producing low-porosity Al2O3/GdAlO3 composites, which present similar mechanical properties to those of alumina. Both the material density and Σ increase with increasing the starting amount of Gd2O3, in a way that for the eutectic composite the density is 50% higher than that of pure alumina, and Σ is about 550000 c.u. The effect of the material porosity on Σ was also considered. The obtained results indicate that the proposed processing route is very attractive for producing these ceramic composites for technological applications.

Calculation of double differential neutron cross-sections of 56Fe and 90Zr isotopes

This study is concerned with the calculations of double differential neutron cross-sections of the structural fusion materials of 56Fe and 90Zr isotopes that are bombarded with protons. Calculations were performed using the level density models of the TALYS 1.95 code and PHITS 3.22 Monte Carlo code. Constant Temperature Fermi Gas, Back Shifted Fermi Gas, and Generalized Super Fluid Models were employed for level density models. Calculations were performed at 22.2 MeV proton energies. Calculations were compared with the experimental data taken from Experimental Nuclear Reaction Data (EXFOR). In conclusion, the results showed that the level density model results of TALYS 1.95 codes for the double differential neutron cross-sections of 56Fe and 90Zr isotopes are consistent with experimental data. On the other hand, PHITS 3.22 results gave lower cross-section values than experimental data at 120 and 150°.

Synthesis, optical characteristics, and gamma-ray protection competence evaluation of polyvinyl chloride reinforced basalt nanocomposites

In this study, new PVC (polyvinyl chloride) polymers doped with different contents of basalt nanoparticles have been prepared. The PVC was mixed with concentrations (0, 5, 10, 15, 20, and 50 wt%) of basalt nanoparticles utilizing the casting method. The samples coded as PB0.0, PB5.0, PB10.0, PB15.0, PB20.0, and PB50.0. The density of the casted samples increased from 1.38 to 2.18 g/cm3 with increasing basalt nanoparticle concentrations. Scanning electron microscope (SEM) of the prepared samples showed a good dispersion with small basalt concentrations in the while some agglomerations were found with the highest basalt concentrations in the mixture. The values of the direct optical energy gap varied from 3.0 to 3.5 eV, while the refractive index values changed from 2.39 to 2.27. The gamma-ray shielding parameters were evaluated using the Phy-X/PSD software program as well as the MCNP – code. The most critical results showed that the linear-attenuation coefficient (μ) increased from 14.421 to 23.862 cm−1 (at 0.015 MeV), and from 0.030 to 0.048 cm−1 (at 15 MeV) for PB0.0 and PB50.0 samples, respectively. Moreover, the evaluation of other gamma - ray shielding characteristics showed an enhancement with the addition of the basalt nanoparticles. Besides, neutron cross section was also calculated, and results revealed that PB0.0 is 0.086 cm−1 and PB50.0 is 0.104 cm−1. Finally, the obtained results concluded that the γ -ray, and neutron shielding capacities were enhanced with the increase in basalt nanocomposites amount in the fabricated composites.

Exploring the potential of BBNCo glasses: Physical, optical, and radiation shielding analysis

The ability of new Co3O4-doped bismuth borate glasses to block radiation was investigated in this study. The glasses had compositions of 55B2O3–35Bi2O3-(10-x)Na2O-xCo3O4(where x = 0, 0.75, 1.5, 2.25, and 3 wt%). The optical properties of the prepared bismuth borate glass doped with Co3O4ions have been studied. The UV demonstrates that as the Co3O4substitution ratio was increased, the samples' absorbance increased noticeably. The optical band gap energies decreased with increasing Co3O4concentration in the case of direct transition and in-direct transition. In addition to the Urbach energy EUvalues of the prepared glass samples was decreased as the concentration of Co3O4ions increased. Using the Phy-X/PSD program, an investigation was conducted on the gamma-radiation shielding capabilities of new BBNCo glasses. The GLAC (linear attenuation coefficient) readings for BBNCo increased from a previous value of 0.7575 cm2/g to a new value of 0.8589 cm2/g when the energy was 356 keV. In addition, computations have been made to determine the removal cross-sections of fast neutrons and the neutron half value layer (NHVL). The addition of Co3O4seems to increase the efficiency of bismuth borate glass in deflecting gamma and neutron radiation, as shown by the findings.

Recent Progress in Gd-Containing Materials for Neutron Shielding Applications: A Review.

With the rising demand for nuclear energy, the storage/transportation of radioactive nuclear by-products are critical safety issues for humans and the environment. These by-products are closely related to various nuclear radiations. In particular, neutron radiation requires specific protection by neutron shielding materials due to its high penetrating ability to cause irradiation damage. Herein, a basic overview of neutron shielding is presented. Since gadolinium (Gd) has the largest thermal neutron capture cross-section among various neutron absorbing elements, it is an ideal neutron absorber for shielding applications. In the last two decades, there have been many newly developed Gd-containing (i.e., inorganic nonmetallic-based, polymer-based, and metallic-based) shielding materials developed to attenuate and absorb the incident neutrons. On this basis, we present a comprehensive review of the design, processing methods, microstructure characteristics, mechanical properties, and neutron shielding performance of these materials in each category. Furthermore, current challenges for the development and application of shielding materials are discussed. Finally, the potential research directions are highlighted in this rapidly developing field.

Neutron attenuation features and elastic properties of silicate glasses containing Ta2O5, and Li2O

In this investigation, the elastic properties and neutrons attenuation factors for some optical glasses containing Ta2O5, SiO2, and Li2O were reported. The present glasses were also consisted of ZrO2 and Nb2O5 in very small concentrations. The glasses are chemically defined as 26.47Li2O-5.88ZrO2-(20-x)Ta2O5-xNb2O5-47.06SiO2, where, x takes the values: 0, 2.94, 5.88, and 11.77 mol%. The elastic properties of these glassy specimens were determined by employing Makishima-Mackenzie's theory (M.M.T). By using the same method, moreover, the micro-hardness and Poisson's ratio were assessed. Cross sections for slow, moderated, and fissile neutrons were computed through standard expressions and models. In addition, the influence of the partial replacement of Ta2O5 by Nb2O5 on the parameters were also analysed. The glass with the lowest Nb2O5 content presented the highest cross sections for fast, moderated, and slow neutrons. The neutron-absorption ability of included glasses declined as glass density declined and Nb2O5 molar concentration increased in the glasses. Therefore, the sample with the highest Ta2O5 content is recommended for neutron absorption applications.

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.

Synthesis and Characterization of Zinc-Lead-Phosphate Glasses Doped with Europium for Radiation Shielding

Appropriate glass systems can provide efficient transparent radiation shielding. The current study involved the preparation of the glass system with a composition of xEu2O3-(15-x)ZnO-10CaO-35PbO-40P2O5 (where x = 0, 1, 2, 3, and 4 wt.% Eu2O3). The formation of the glass phase was confirmed using X-ray diffraction (XRD). The study analyzed physical and structural parameters, such as optical conductivity (σopt), refractive index (n), and optical band gap (Eg), with the amount of Eu2O3. The findings indicate that the optical band gap increased as the Eu2O3 content increased. Additionally, a decrease in Urbach energy (EU) was observed, suggesting an improvement in the orderliness of the glass. The study also determined various parameters for gamma-ray shielding, including mass attenuation coefficient (μm), effective atomic number (Zeff), and kerma coefficient (k). For neutron shielding characteristics, the macroscopic effective removal cross-section (ΣR) of fast neutrons was calculated.

Radiative neutron capture reaction rates for nucleosynthesis: The creation of the first r-process peak

About half of the elements beyond iron are synthesized in stars by rapid-neutron capture process (r-process). The stellar environment provides very high neutron flux in a short time ($\sim$ seconds) which is conducive for the creation of progressively neutron-rich nuclei till the waiting point is reached after which no further neutron capture reactions proceed. At this point such extremely neutron-rich nuclei become stable via $\beta^-$ decay. A detailed understanding of the r-process remains illusive. In the present work, we explore the radiative neutron-capture (n,$\gamma$) cross sections and reaction rates around the r-process peak near mass number eighty. The inherent uncertainties remain large in some cases, particularly in case of neutron-rich nuclei. When the low-energy enhancement exists, it results in significant increase in the reaction rate for neutron-capture.