Neutron Absorption Cross Section Research Articles

Binder jetting and melt infiltration for ceramic-metal fuel fabrication

This study explored the feasibility of fabricating a ZrO2-Zr2Cu ceramic-metal (CERMET) fuel through additive manufacturing (AM) followed by melt infiltration. ZrO2 was selected as a surrogate for UO2 owing to its similar properties, and Zr2Cu was selected owing to its low melting point and neutron absorption cross-section to explore the feasibility of forming a uniform distribution of ceramic particles within a metal matrix. The binder jetting (BJ) AM technique was employed to fabricate preforms with precise control over the particle distribution, addressing the challenges associated with particle agglomeration and ensuring preform shape integrity. Subsequently, spontaneous melt infiltration was employed to achieve a dense and homogeneous CERMET structure. This study emphasized the significance of controlling wettability and surface energies for successful melt infiltration, focusing on the interfacial reactions between Zr2Cu and ZrO2. These results indicate that this novel approach can significantly enhance the thermal and irradiation performance of CERMET fuels.

The hydrogen-helium-vacancy interaction and hydrogen-vacancy clusters formation mechanisms in chromium: A first-principles study

Attributing to excellent corrosion resistance and low neutron absorption cross-section, chromium (Cr) has been considered as a promising advanced structural nuclear material. Under extreme radiation conditions, the coexistence of vacancies (Vac), hydrogen (H), and helium (He) modifies the evolution trajectory of H-defect clusters, and significantly impacts the performance of materials, all of which are under extensive investigations. Here, the occupancy mechanism of H atoms and influencing factors in Vac1/He1Vac1 systems are comprehensively studied by first-principles calculations. Based on the exhaustive investigation of H-He-Vac interactions, the equilibrium distance between H-H pairs is delineated. Serving as an effective trap center, the Vac1/He1Vac1 cluster exhibits robust attractive forces capable of luring dissociative H atoms within the crystal lattice. The accumulation of H atoms from remote regions towards the defects results in the formation of stable Hn-Vac1/Hn-He1Vac1 complexes. Moreover, it is revealed that the hydrogen-enhanced vacancy mechanism and meticulously examines the factors affecting defect-trapping capabilities and the growth of H-He-Vac complexes, providing an insight into the behavior of hydrogen-defects clusters formation mechanism and the H-He-Vac interaction in Cr.

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.

Analysis of the atomic ratio of H and Zr effect on the neutronics parameters of ZrH moderated space nuclear reactor

Zirconium hydride (ZrH) is an ideal moderator for space nuclear reactors due to its exceptional properties, including high hydrogen content, small neutron absorption cross section, and high operating temperature. The primary goal of this study is to examine how the atomic ratio of H to Zr (H/Zr ratio) influences the neutronics parameters of a ZrH moderated space nuclear reactor, with a focus on establishing a reliable reference for ensuring the optimal safety and minimization of such reactors. The neutronics calculation based on the ZrH moderated space nuclear reactor named Topaz-II is performed using the Reactor Monte Carlo code (RMC code) with the ENDF/VII cross-section database. The effects of the H/Zr ratio were studied with a particular focus on the initial keff, the burnup, the temperature reactivity coefficient and the criticality safety. The results show that with the increase of the H/Zr ratio, the initial keff increases while the drums’ worth decreases. The Moderator Temperature Coefficient (MTC) is a positive value that rises as the H/Zr ratio increases. In the dropping accidents, the reactor full of voids with seawater is more serious, and the introduced reactivity decreases with the increase of the H/Zr ratio.

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.

Accident tolerant fuel design optimization for SMART reactor

Fully Ceramic Microencapsulated (FCM) fuel is one of the most advancedfuel types for the next nuclear reactor generations. The use of uranium carbide fuel in FCM offers multiple benefits such as high fuel density, improved thermal conductivity and enhanced mechanical stability. Both UC and conventional fuel (UO2) have the same melting point. As a result, under reactor operating conditions, UC has better safety margins. In addition, FeCrAl is a promising cladding material. However, because of its greater thermal neutron absorption cross section, FeCrAl has a neutronic penalty when compared to Zircaloy leading to an annual reload requirement with high fuel enrichment. Owing to their low power density and excellent design, small modular reactors (SMRs) have exceptional inherent safety. In this work, we determined the optimum design of an accident tolerant fuel (ATF), with respect to neutronics, in SMART reactor using dense UC FCM (in contrast to the conventional UO2 fuel) and a combination of zircalloy and FeCrAl (FCM -Zircaloy, FCM -FeCrAl, FCM -combining of zircaloy and FeCrAl) as cladding to achieve better thermal conductivity as well as safer performance, while considering the required fuel enrichment for the proposed designs. The effective multiplication factor and burnup were calculated for neutronic assessment of the SMART core, the outcomes are showcased alongside a comprehensive assessment of the characteristics and prospective. The neutronic parameters were estimated using the stochastic MCNPX 2.6 code. We found that Fully Ceramic Microencapsulated -Uranium Carbide (FCM-UC) fuel type was more attractive than conventional UO2. The optimum cladding design as compared to the reference case and the other cases of (FCM-UC), was determined. When 50% FeCrAl cladding was employed, case 15 in our study emerged as the superior design, demonstrating maximum production of 239Pu due to higher initial concentrations of 235U and quantities of 238U. This work may offer recommendations for future optimized designs of the SMART reactor parameters.

A comprehensive research on γ-ray and particle radiation interaction parameters of Cerium doped heavy oxide inorganic scintillators

In this study, the interaction cross sections of heavy oxide scintillators, which have a very important place in high energy physics research, with photon and particle radiation were comparatively investigated. In this context, the interaction parameters of six different Ce-doped heavy oxide scintillators were obtained using updated software. The interactions of photons with energies between 1 keV and 100 GeV with scintillators were evaluated in the light of the results produced by the Phy-X/PSD software. In addition to the continuous energy range, calculations were also made on the photon energies emitted with the highest probability of emission from sources 241Am (59.54 keV), 133Ba (356 keV), 22Na (511 and 1280 keV), 137Cs (662 keV) and 152Eu (1460 keV). Additionally, evaluations of energy absorption build-up factors valid under broad beam conditions are also included in this study. Fast neutron absorption cross sections and changes in the range (R) values for heavy particle radiation in the range of 1 keV–20 MeV were also determined with SRIM and ESTAR software. All data obtained depending on energy are explained according to the mass percentages of Lu, Gd, Hf, Y, Al and O elements in the scintillators. The highest interaction cross section for photons was obtained in the Lu2Gd2SiO5 (Ce 0.3%) scintillator (HOS2). The interaction cross section of this scintillator was obtained as 3.115 and 0.042 cm2/g at 0.1 MeV and 10 MeV photon energies, respectively. On the other hand, the highest cross-sections of the fast neutrons were observed in the scintillator (HOS5) containing GdAlO3 (Ce 1.0%) as 0.148 cm−1. This difference in neutron radiation has been explained by calculating the individual attenuation abilities of the elements in the structures. It is hoped that the data obtained will guide researchers working on the detection of different types of radiation.

Implementation of thorium-based fuel for Indonesia Micro Reactor (IMR)

Indonesia has committed to reducing greenhouse gas emissions in the framework of net zero emissions, and nuclear energy will be a part of the energy mix. Indonesia has been developing a 10 MWt micro reactor of high temperature gas-cooled reactor (HTGR) with pebble fuel containing 17 % enriched UO2 kernel with OTTO cycle scheme. For the government’s interest to use thorium, this paper assesses the possibility of replacing uranium with thorium. The assessment is performed by calculating the effective neutron multiplication factor (keff). Neutrons captured by 232Th lead to the formation of two intermediate isotopes, namely 233Th and 233Pa, the latter being a significant neutron absorption cross-section. Neutronics analysis is performed using parameters of thorium mass fraction, core temperature, and active core height. The thorium mass fractions varies from 0 % to 90 % Th. The core temperatures are in the range of the normal operating temperature at 900 K, assumed superheat temperature at 1200 K, and shutdown temperature at 300 K. The active core height varies at 197, 180, 150, 130, and 125 cm. This paper aims to determine the keff of the IMR at maximum thorium/uranium mass fraction composition, maximum active core height, core temperature, and fuel burn-up. The MCNP code is used for analysis. The results show that keff decreases with increasing thorium mass fraction and core temperature, and decreasing active core height. Calculations with thorium mass fraction of 15 %, core active height of 197 cm in 20 steps show that the reactor may be continuously operated for more than 920 days without refueling. The analysis results show that IMR design has negative temperature reactivity, inherent safety characteristics, and the ability to use thorium-based fuels.

Microstructure and Properties of Pressureless-Sintered Zirconium Carbide Ceramics with MoSi2 Addition.

Zirconium carbide (ZrC) ceramics have a high melting point, low neutron absorption cross section, and excellent resistance to the impact of fission products and are considered to be one of the best candidate materials for fourth-generation nuclear energy systems. ZrC ceramics with a high relative density of 99.1% were successfully prepared via pressureless sintering using a small amount of MoSi2 as an additive. The influence of the MoSi2 content on the densification behavior, microstructure, mechanical properties, and thermal properties of ZrC ceramics was systematically investigated. The results show that the densification of ZrC was significantly enhanced by the introduction of MoSi2 due to the formation of a liquid phase during sintering. In addition, the ZrC grains were refined due to the pinning effect of the generated silicon carbide. The flexural strength and Vickers hardness of ZrC ceramics with 2.5 vol% MoSi2 sintered at 1850 °C were 408 ± 12 MPa and 17.1 GPa, respectively, which were approximately 30% and 10% higher compared to the samples without the addition of MoSi2. The improved mechanical properties were mainly attributed to the high relative density (99.1%) and refined microstructure.

Hot deformation behaviour of Swaged Zircaloy-4 at varying strain rate and its microstructural evolution

Zirconium alloy demonstrates a favourable combination of strength, ductility, low neutron absorption cross-section, as well as excellent corrosion and oxidation resistance. In the context of its application as a reactor component, the critical hot deformation behaviour of isothermally swaged Zr-4 alloy has been extensively studied. Different combinations of elevated temperatures (873 K to 1023 K) and strain rates (0.01 s−1 to 10/s) have shown enhancement in flow stress. Analysis of stress flow, microstructural changes, and activation energy behaviour was performed using a state-of-the-art thermomechanical simulator with 50% reduction in sample thickness. The modified Arrhenius equation was employed to correlate theoretical and experimental flow stress behaviour. EBSD was utilized to understand texture evolution in hot deformed samples, revealing a transformation from pure basal to basal plus pyramidal poles concerning the compression direction. Texture evolution played a significant role in microstructural development, further characterized through TEM, which showed influential characteristics like subgrain boundaries, dislocations, and precipitate behaviour. Dynamic recrystallization mechanisms in swaged Zr-4 hot deformed samples were analysed using bright and dark field TEM modes, which were subsequently correlated with other microstructural characteristics such as grain morphology and structural defects.

A novel aloe vera mediated Bismuth–Magnesium–Copper nanocomposite for radiation shielding applications

In the current study, Bismuth-Magnesium-Copper nanocomposites (BMC NCs) are synthesized using versatile solution combustion synthesis using Alovera extract as a reducing agent. As obtained BMC NCs are characterized using standard techniques. Bragg’s reflections confirmed the formation of multi-phase BMC NCs with crystal sizes 18.11 nm and 21.12 nm. The surface morphology of BMC NCs confirmed the agglomeration of spherical-shaped nanoparticles. EDAX spectrum analyzed the presence of Bi, Mg, and Cu. The direct energy gap is found to be 3.08 eV. X-ray/gamma radiation shielding properties are then evaluated using a NaI(Tl) detector coupled with MCA. Shielding parameters are slightly higher than lead. Neutron shielding parameters are compared with conventional shielding materials. Scattering length is found to be 8.8 fm, and neutron absorption cross-section and neutron attenuation parameters of BMC NCs are found to be efficient when compared with lead, concrete, and steel. Bremsstrahlung radiation shielding parameters are computed at 1.511 MeV. Bremsstrahlung efficiency is comparable to lead, bremsstrahlung dose rate and specific bremsstrahlung constant are found to be more in comparison with lead. Thus, BMC NCs are highly efficient in shielding X/gamma ray and neutron radiations.

Tribological and corrosion behavior of hydrided zirconium alloy

Zirconium alloys (Zr-alloys) are used for the core components of nuclear reactors due to their exceptional properties, which make them suitable for the given applications such as low neutron absorption cross-section, good corrosion resistance, good wear resistance, good mechanical stability at high temperatures etc. The coolant water that flows inside the pressure tube contains LiOH to control the pH value of the coolant water, which can cause corrosion. The flow of coolant water produces low amplitude vibration, which can cause fretting wear of the components. Further, the coolant water at high temperature causes the hydride formation in the material, degrading the properties of zirconium alloys. The safe working of nuclear reactors is the prime objective, and the corrosion and wear behavior of the core components must be explored to prevent possible failure. Fretting wear or tribo-corrosion experiments were conducted on Zr‐2.5Nb alloy against SS-410 for varied fretting duration to understand the tribological response. Electrochemical corrosion experiments were performed under the aqueous solution of LiOH (10.4 pH) using Electrochemical Impedance Spectroscopy (EIS) and potentiodynamic polarization technique. The comparison was made between the tribological and corrosion response of unhydrided‐Zr‐2.5Nb alloy (Zr‐2.5Nb) and hydrided- Zr‐2.5Nb alloy (Zr‐2.5Nb‐H). The hydride formation causes the reduction in corrosion resistance of Zr‐2.5Nb alloy. The oxide layer is formed during the wear and corrosion process which protects the material from further wear and corrosion respectively.

Monte Carlo simulation and optimization of neutron ray shielding performance of related materials

Polymers have become widely used substrate materials for shielding neutron rays because of their high hydrogen content and easy processing procedures. Rare-earth materials are also being gradually adopted as neutron absorbers because of their considerable thermal neutron absorption cross-sections. This paper utilizes the FLUKA Monte Carlo simulation program to compare the shielding effects of various polymers and rare-earth oxides on neutron rays across different energy ranges. The study investigates the superior shielding materials for neutron radiation in each energy range. Subsequently, a series of materials are simulated by combining the preferred shielding materials for neutron rays in each energy range, exploring the influence of material composition and composite structure on the effectiveness of neutron ray shielding. It is revealed that the preferred material for shielding neutron rays changes for different energy ranges. For low-energy neutron rays, rare-earth oxides such as Sm2O3 and Gd2O3 demonstrate the most effective shielding, whereas for high-energy neutron rays, polyethylene (PE) provides the best shielding performance. Materials with different compositions show varying preferred structures when dealing with a 252Cf neutron source. However, in mitigating the secondary gamma rays generated during the neutron shielding process, stacked-type materials exhibit the most effective shielding performance.

Determination of 10B/11B in boric acid and B4C using LA-ICPMS

Abstract Boron isotopic ratio measurement is very important in the nuclear industry because it is used as neutron poison (in the form of boric acid or a B4C pellet) and control rod material (B4C pellet) in nuclear power plants. Since 10B has a higher neutron absorption cross-section, enriched 10B is used in the nuclear industry and the extent of enrichment varies from 50–90 %. Therefore, it is essential to determine the 10B/11B ratio in boric acid and B4C. The prime purpose of the present study is to utilize LA-ICPMS for evaluating 10B/11B ratio both in solution (boric acid) and solid (B4C pellet) without any pre-treatment so that the difficulty in dissolution or powdering process of B4C (one of the hardest materials) can be avoided and the B4C pellet can be reused. The results obtained from LA-ICPMS for the B4C pellet are discussed and compared with the solution ICPMS, TIMS, and LIMS. The solutions of boric acid (natural abundance) and dissolved B4C (natural & ∼67 % enriched isotopic composition) were analyzed by LA-ICPMS and validated by liquid sampling ICPMS data.

Effectiveness of green-synthesized nickel-doped calcium ferrite nanoparticles in the X-ray/gamma radiation shielding applications

Nanoparticles comprising calcium ferrite doped with Nickel within a concentration range of 10 to 50 mol% were successfully produced using a solution combustion technique, employing neem leaf extract as a reducing agent. The subsequent calcination process at 500 °C validated the formation of orthorhombic calcium ferrite, featuring distinct Bragg reflections indicative of Nickel doping. An analysis via scanning electron microscopy (SEM) affirmed the presence of irregularly shaped nanoparticles with pores and voids within their structure. The optical energy band gap, determined through Wood and Tauc’s relation, ranged from 2.89 to 2.97 eV. Notably, an increase in Nickel substitution was found to correlate with an increase in the energy band gap and a reduction in crystallite size. The research extensively evaluated the nanoparticles’ efficacy in radiation protection, encompassing parameters such as mass attenuation coefficient, mean free path, half-value layer, and effective atomic number. The mass attenuation coefficient is found to be 0.05353 cm2g−1 for 40 mol% at 1332 KeV. EABF is found to be large for 40 mol% at 0.89 MeV energy. Likewise, 50 mol% has a high Neutron absorption cross section (σab) which is 1.5(barn). Also, the bremsstrahlung dose rate is found to be 1.35 (R/h)*103, at 1.511 MeV.Additionally, the study comprehensively explored variations in the energy-absorbing buildup factor across diverse energy ranges, taking into account the Nickel concentration within the calcium ferrite nanoparticles. This aspect holds significant importance in the field of radiation dosimetry, particularly in applications related to neutron shielding and bremsstrahlung radiation.

Phase-field simulation of high-temperature corrosion of binary Zr-2.5Sn alloy

Due to the small neutron absorption cross section and excellent thermal creep performance, zirconium alloy is one of the most important cladding materials for fuel rods in commercial fission reactors. However, quantitative analysis of the effects of temperature and grain boundaries on the corrosion microstructure evolution of zirconium alloys is still needed. The establishing of a phase field simulation for the corrosion process of polycrystalline zirconium alloy and the systematical investigating of the thermodynamic influence are both very important. In this study, the phase field model of the corrosion process in zirconium alloys is developed by combining corrosion electrochemistry through calculating the interfacial energy at the metal-oxide and oxide-fluid boundaries. Then the model is used to investigate the uniform corrosion behavior on the surface of Zr-2.5Sn alloy, which demonstrates that the corrosion kinetic curve follows a cubic rule. Subsequently, the influence of temperature on the corrosion thickening curve of zirconium alloy is examined, and good agreement between simulation and experimental results is achieved. It is observed that during early stage of oxide layer formation, there is a high growth rate with minimal temperature dependence; however, as the oxide layer thickness increases, temperature becomes a significant factor affecting its growth rate, with higher temperatures resulting in faster corrosion rates. Furthermore, the effect of polycrystalline zirconium alloy matrices on corrosion rate is investigated, revealing that the grain boundaries accelerate oxide layer thickening due to enhanced oxygen diffusion rates. At metal-oxide interface, O<sup>2–</sup> bands are formed in areas with higher O<sup>2–</sup> concentration along these grain boundaries towards the metal matrix, which mainly influences oxidation-corrosion rate during the initial oxidation stage.

Study on damage of Gd2O3-CeO2 under electronic energy loss: comparison between bulk-like and nanostructure.

To understand the physical phenomena responsible for radiation damage of the materials used in nuclear reactors, and thus study their operation life and/or efficiency, it is required to simulate the conditions by exposing the materials to energetic ions. Ceria (CeO2) has been proposed as one of the inert matrices for the transmutation of minor actinides in the futuristic inert matrix fuel (IMF) concept. The inert matrix should also contain burnable poison to compensate for the initial reactivity of fuel. In this context, gadolinium (Gd) is an excellent burnable poison with a high neutron absorption cross-section. In view of this, Gd2O3-CeO2 nano-powders were synthesized and sintered at 800 °C and 1300 °C to obtain different grain sizes and morphologies. FESEM and TEM were carried out to study the grain size of pristine pellets. The sintered pellets were irradiated with 80-MeV Ag ions (electronic energy loss (Se) regime) at room temperature to emulate the effect of fission fragments. For analysis of the effect of grain size on the irradiation-induced structural degradation at different fluences, GIXRD and Raman spectroscopy were performed. Significantly large damage has been observed for the smaller grain-sized samples (sintered at 800 °C) as compared to the large grain-sized sample (sintered at 1300 °C). Neither of the samples amorphized under the present experimental conditions as indicated by the presence of the Raman-active T2g mode (centred at 462 cm-1) and all the XRD peaks of fluorite cubic structure up to the highest fluence employed (1 × 1014 ions cm-2). X-ray photoelectron spectroscopy results demonstrate that Ce4+ to Ce3+ and vacancy-related isolated clusters are the main defects produced in the systems. The radiation tolerance behaviour of the samples is understood with the help of thermal spike simulation, which indicates higher transient lattice temperatures with longer duration in the smaller grain-sized sample upon irradiation. Gd-doped ceria thus possesses good radiation stability in the Se regime, indicating its potential for application in IMFs.

Neutron shielding analysis for a gadolinium doped nickel alloy

Rare-earth-nickel (RE-Ni) alloy is a kind of appealing neutron shielding material for advanced reactors and spent fuel storage containers. The incorporation of neutron-absorbing additive RE elements results in a RE-Ni alloy with integrated structural and functional features, which can drastically reduce shielding material weight. Gadolinium (Gd), in comparison to other rare-earth elements, has a much larger thermal neutron absorption cross section, necessitating only a low fraction to achieve the desired shielding capability. Hence, gadolinium is doped into a nickel-based alloy to form a gadolinium-nickel (Gd-Ni) alloy. Due to its extremely low solubility in the nickel phase, gadolinium is commonly manifested as heterogeneously spatially scattered Gd-rich phases in existing preparation procedures of alloys rather than homogeneously distributed gadolinium elements. This would result in a strong neutron tunneling effect, limiting the neutron shielding capability of Gd-Ni alloys. In this work, a superfine layered Gd-Ni alloy doped with Gd-rich phases and arranged in dislocations is proposed to provide an adequate dislocation density of Gd-rich phases to suppress the neutron tunneling effect. With the same Gd content, the thermal neutron shielding ability of a layered Gd-Ni alloy is comparable to the case that Gd is assumed to be an ideally homogeneous distribution in the material. The smaller the particle size of the Gd-rich phase, the weaker the neutron tunneling effect, and the lower the thermal neutron transmission ratio. For a 0.5 cm-thick Gd-Ni alloy, the particle size of Gd-rich phase should be kept within 20μm to achieve a thermal neutron transmission ratio of less than 10−5 for natural gadolinium content of 0.6 at.%. This work provides an important guidance for the design and fabrication of the next-generation neutron shielding materials.

Density functional study on the mechanics, thermodynamics, and H diffusion mechanism of LiH

LiH is valued by researchers and engineers for its high-temperature resistance, high strength, high hydrogen storage capacity of up to 12.6 wt%, and high neutron absorption cross section. However, sintering and application of this material are difficult. Lithium hydride is a highly effective neutron shielding material. The neutron shielding ability of lithium hydride is directly affected by the homogeneity of the total amount and distribution of hydrogen. The heat migration of hydrogen reduces the neutron shielding effectiveness of lithium hydride. It is critical to forecast H's diffusion behavior. The diffusion coefficient is a critical metric for determining the diffusion efficiency. As a result, the stability, electronic structure, and bonding properties of LiH, LiOH, and H entering the tetrahedral interstitial sites of LiH were calculated using first principles based on density functional theory (DFT). The elastic modulus of the system was also examined to find the lowest thermal conductivity, and some thermodynamic parameters were derived by calculating LiH phonons. Finally, static calculations were used to determine the diffusion coefficient and diffusion activation energy of H in the LiH solid solution. The results revealed that the addition of H altered the characteristics of the LiH semiconductor, giving it metallic features. The lowest thermal conductivity of the LiH system was 2.9. W·m−1 K−1. The diffusion coefficient was D=(2.638×10−2cm2/s)exp(−5.811×10−3cal/KBT). The activation energy for diffusion was 0.252 eV. The diffusion coefficients, elastic constants, and lattice parameters all corresponded well with previously published results. The density functional theory was employed to study the material's system energy and electrical structure, and the microscopic migratory behavior of hydrogen was predicted theoretically. The calculation of the lowest thermal conductivity provided a reference for the screening of ceramic materials.

Numerical analysis of multi-field coupling behaviors of delayed hydride cracking for thin plate-type Zr-2.5Nb specimens

Zirconium alloys are extensively utilized as structural materials in nuclear applications due to their exceptional mechanical strength, corrosion resistance, and low thermal neutron absorption cross-section. The stress concentration provoked by the defects in zirconium alloy components will initiate delayed hydride cracking (DHC), significantly jeopardizing nuclear safety. In the present investigation, an enhanced cohesive model that incorporates the effects of stress state has been proposed to realize the three-dimensional simulation of multi-field coupled DHC behaviors in thin plate-type zirconium alloy specimens. The simulation results indicate that: (1) the consistence of the predicted DHC velocities with the corresponding experimental data validates the newly developed cohesive model; (2) the strip characteristics presented in the cracked surface are captured, demonstrating that the varying fracture patterns across the thickness are well simulated; (3) the stress-state induced hydrostatic-stress gradients and solid-soluted hydrogen concentration gradients make against the formation of a hydride strip to induce hydride embrittlement near the plate surfaces, resulting in a slower cracking of the surface layers than that of the internal region; for the thinner specimens the asynchronous cracking phenomena through the thickness will greatly affect the multi-field coupling behaviors around the mid-plane; the non-cracked surface layers will apply the dragging effects to the new cracking of the internal region, leading to longer waiting time and thicker hydrided zone before fracture.