Neutron Absorption Cross Section Research Articles

Temperature-dependent lattice dynamics in iridium

The characterization of simple elemental systems is key to benchmarking first-principles modeling of electronic and vibrational behaviors of materials. A large body of literature has been built for most elemental systems in the periodic table. However, surprisingly little neutron work has been performed to probe the vibrational properties of iridium, likely due to its large neutron absorption cross section. Nonetheless, iridium is of significant scientific and technological interest due to large relativistic electron effects and electron-phonon coupling, particularly in strongly correlated iridate compounds. Here, we report temperature-dependent inelastic neutron scattering measurements of the vibrational properties of iridium, from which we extract key thermodynamic properties. To overcome the challenge of the large neutron absorption of iridium, we developed a simple postprocessing correction procedure. The measured densities of phonon states compare well with quasiharmonic density functional theory calculations, although the obtained experimental phonon Gr\uneisen parameters are much larger than expected, reaching as high as $\ensuremath{\gamma}=4.5$, indicating substantial anharmonicity.

Boron-Based Neutron Scintillator Screens for Neutron Imaging.

In digital neutron imaging, the neutron scintillator screen is a limiting factor of spatial resolution and neutron capture efficiency and must be improved to enhance the capabilities of digital neutron imaging systems. Commonly used neutron scintillators are based on 6LiF and gadolinium oxysulfide neutron converters. This work explores boron-based neutron scintillators because 10B has a neutron absorption cross-section four times greater than 6Li, less energetic daughter products than Gd and 6Li, and lower γ-ray sensitivity than Gd. These factors all suggest that, although borated neutron scintillators may not produce as much light as 6Li-based screens, they may offer improved neutron statistics and spatial resolution. This work conducts a parametric study to determine the effects of various boron neutron converters, scintillator and converter particle sizes, converter-to-scintillator mix ratio, substrate materials, and sensor construction on image quality. The best performing boron-based scintillator screens demonstrated an improvement in neutron detection efficiency when compared with a common 6LiF/ZnS scintillator, with a 125% increase in thermal neutron detection efficiency and 67% increase in epithermal neutron detection efficiency. The spatial resolution of high-resolution borated scintillators was measured, and the neutron tomography of a test object was successfully performed using some of the boron-based screens that exhibited the highest spatial resolution. For some applications, boron-based scintillators can be utilized to increase the performance of a digital neutron imaging system by reducing acquisition times and improving neutron statistics.

Continuous-energy Monte Carlo calculation analysis of light-water moderation critical experiments with aqueous solution samples of B, Rh, Cs, Nd, Sm, Eu, Gd, and Er

ABSTRACT In order to validate the neutron absorption cross sections in the JENDL-4.0 nuclear data library, critical analysis was performed using a continuous-energy Monte Carlo calculation code for the light-water moderation UO2 cores loaded with the aqueous solution samples of eight elements (B, Rh, Cs, Nd, Sm, Eu, Gd, and Er). The concentrations of the elements in the solution samples were changed in the experiments, and the criticality water levels were measured. The calculated keff values of the critical cores for the different concentrations of each element were plotted against the reactivity worth of the solution samples calculated based on the differences in the critical water levels. They were fitted by an approximate linear curve to obtain the slope which was regarded as a negative value of a fractional bias of the absorption cross section in the thermal and epi-thermal neutron energy regions. The obtained fractional biases indicated that the absorption cross sections of Sm, Eu, and Gd agreed with the actual values with the uncertainties of about 3%, and those of B, Rh, and Er overestimated them by 6 3%, 5 3%, and 6 4%, respectively.

Evaluation of shielding efficiency of using boric acid as coolant in CFETR vacuum vessel

Vacuum vessel is a hermetically sealed steel container made up of double steel walls and in-wall shielding, which shields the neutron and protects other components in fusion reactor. Cooling water circulates through the vessel's double steel walls and removes the heat generated during operation. Since boron-10 has a large neutron absorption cross-section, boric acid is supposed to have better shield effect than the original cooling water. In this paper, the effectiveness of pure water and boric acid watery solution as coolant between the double steel walls of the vacuum vessel of CFETR is evaluated by Monte Carlo method. In the simulation, shield efficiency of pure water and different concentration of boric acid is investigated. Neutronic characteristics including neutron flux and neutron spectrum have been calculated. The result shows that boric acid can significantly reduce the flux of total neutron, while the flux of fast neutron takes less change. In conclusion, boric acid has better neutron shielding efficiency than water and can make total neutron flux decline 35%.

Finite Element Analysis and Experimental Evaluation of Residual Stress of Zr-4 alloys Processed through Swaging

Zirconium alloy has been extensively used as a cladding material in nuclear power reactors due to its low neutron absorption cross section, excellent mechanical properties, and corrosion resistance. The influence of the swaging parameter, feed rate (0.7, 1.25, 2 m/min) on residual stress induced in Zr-4 alloy is investigated in the present work. A three-dimensional finite element model was implemented in the Deform 3D software to simulate the rotary swaging (RS) process over a circular rod of Zr-4 alloy. The simulation results based on the 3D framework provide a detailed insight of residual stress, true stress versus true strain and force applied over the rod during the multiple pass swaging process; the results are compared with experimental results. The experimental hole drilling method is used to determine the residual stresses on swaged zirconium alloy at different feed rates (0.7, 1.25, and 2 m/min). A similar trend of residual stress between experimental and numerical results from the surface to the center on the swaged rod samples is observed. The same magnitude of residual stress at the surface of the swaged Zr-4 rod is also observed. It is found to be compressive at the surface and tensile in the center of the samples, as observed in the present work.

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

10B-based films grown by pulsed laser deposition for neutron conversion applications

Solid-state neutron detectors exploit nuclear interactions producing energetic charged particles within a converter layer embedding nuclei with high neutron capture cross section and with thickness compatible with the charged particle range. Recently, boron-10 (10B) is being considered as a valid alternative to the expensive and decreasingly available 3He gas thanks to its large neutron absorption cross section and high-energy reaction products. Minimized amounts of impurities, films with optimal and well-controlled thickness, good uniformity over large areas and good adhesion to the substrates are essential to achieve efficient neutron detection performance. In this study, we present well-adherent 1-µm-thick 10B-enriched boron coatings, deposited over a large area by off-axis pulsed laser deposition (PLD) using a nanosecond Nd-YAG laser beam operating at 1064 nm and high fluence (~ 10 J/cm2), onto 1-mm-thick Al substrate as well as smooth and rough C substrates. By combining plasma plume divergence and peaked profile under off-axis deposition geometry, uniform10B films were obtained over an area of a 3.4 × 3.4 cm2. We discuss the morphological characteristics of our deposits as related to the mechanisms of nanosecond laser ablation and present energy dispersive spectroscopy (EDS) elemental analysis of the film, paying attention to the different surface features. Negligible presence of O, N and C contaminants was achieved by controlled vacuum conditions. Moreover, we present encouraging neutron detection performances of our film deposited onto Al.

TeO2–B2O3–ZnO–La2O3 glasses: γ-ray and neutron attenuation characteristics analysis by WinXCOM program, MCNP5, Geant4, and Penelope simulation codes

γ-ray, fast and thermal neutrons shielding features for five Pb-free {[(TeO2)0.70(B2O3)0.30]0.7 (ZnO)0.30}(1-x)(La2O3)x (x = 0.01, 0.02, 0.03, 0.04, and 0.05 mol%) glasses have been inspected in this work. For all samples, within the photon energy range of 0.015–15 MeV, linear attenuation coefficients (μ) and mass attenuation coefficients (μ/ρ) were estimated using WinXCOM program (theoretical approach), and obtained μ/ρ values are in fine agreement with computed μ/ρ results of respective MCNP5, Geant4, and Penelope codes. By utilizing μ/ρ values of WinXCOM, γ-ray attenuation parameters like effective atomic number (Zeff), effective electron density (Neff), half-value layer (HVL), and mean free path (MFP) were assessed within 0.015–15 MeV photon energy range, for all samples. Additionally, exposure buildup factors (EBFs) and energy absorption buildup factors (EABFs) were calculated for all samples by applying a five parameter geometric progression (G‒P) fitting method as a function of various penetration depths (1, 5, 10, 15, 20, 25, 30, 35, and 40 mfp) within 0.015–15 MeV photon energy range. The derived radiation protection efficiency (RPE) results indicate all samples' effective attenuation competence for lower energy γ-rays. Among all glasses, 46.55TeO2-19.95B2O3-28.5ZnO–5La2O3 (mol%) sample, by having relatively bigger μ/ρ and Zeff, minimal HVL and MFP, and lowest EBF and EABF, endorse its' excellent γ-ray shielding effectiveness. Further, neutron shielding parameters such as macroscopic removal cross-section for fast neutrons (ΣR), coherent and incoherent scattering cross-sections (σcs & σics), absorption cross-section (σA), and total cross-section (σT) for thermal neutrons were explored. By using the Geant4 code, ‘σT’ values have been deduced within 10−4‒10−8 MeV neutron energy range. The 46.55TeO2-19.95B2O3-28.5ZnO–5La2O3 (mol%) glass owns relatively larger ΣR (0.1474 cm−1) for fast neutrons attenuation and ‘σT’ (10.1444 cm−1 at 0.0253 eV neutron energy, 15.9824 cm−1 → 0.551 cm−1 from 1 × 10−8 MeV →1 × 10−4 MeV neutron energy) for thermal neutrons absorption, respectively, revealing its' superior neutrons shielding ability.

Realistic Ion Dynamics through Charge Renormalization in Nonaqueous Electrolytes.

While many practically important electrolytes contain lithium ions, interactions of these ions are particularly difficult to probe experimentally because of their small X-ray and neutron scattering cross sections and large neutron absorption cross sections. Molecular dynamics (MD) is a powerful tool for understanding the properties of nonaqueous electrolyte solutions from the atomic level, but the accuracy of this computational method crucially depends on the physics built into the classical force field. Here, we demonstrate that several force fields for lithium bistriflimide (LiTFSI) in acetonitrile yield a solution structure that is consistent with the neutron scattering experiments, yet these models produce dramatically different ion dynamics in solution. Such glaring discrepancies indicate that inadequate representation of long-range interactions leads to excessive ionic association and ion-pair clustering. We show that reasonable agreement with the experimental observations can be achieved by renormalization of the ion charges using a "titration" method suggested herewith. This simple modification produces realistic concentration dependencies for ionic diffusion and conductivity in <2 M solutions, without loss in quality for simulation of the structure.

Safety features of fast reactor with heavy atomic weight weakly neutron absorbing reflector

The purpose of the present study is the justification of the possibility of improving fast reactor safety by surrounding reactor cores with reflectors made of material with special neutron physics properties. Such properties of 208Pb lead isotope as heavy atomic weight, small neutron absorption cross section, and high inelastic scattering threshold result in certain peculiarities in neutron kinetics of the fast reactor equipped with 208Pb reflector, which can significantly enhance reactor safety. The reflector will also make possible generation of additional delayed neutrons characterized by the “dead” time. This will improve the resistibility of the fission chain reaction to stepwise reactivity excursions and exclude prompt supercriticality. Let us note that generation of additional delayed neutrons can be shaped by reactor designers. The relevance of the study amounts to the fact that generation of additional delayed neutrons in the reflector will make it possible mitigating the consequences of a reactivity accident even if the introduced reactivity exceeds the effective fraction of delayed neutrons. At the same time, the role of the fraction of delayed neutrons as the maximum permissible reactivity for reactor safety is depreciated. Scientific originality of the study pertains to the fact that the problem of yield of additional neutrons with properties close to normal delayed neutrons, has not been posed before. The authors suggest a new method for enhancing safety of fast reactors by increasing the fraction of delayed neutrons due to the time delay of prompt neutrons during their transfer in the reflector. In order to benefit from the expected advantages, the following combination is acceptable: lead enriched by 208Pb is used as a neutron reflector while natural lead or other material (sodium, etc.) is used as a coolant in the reactor core.

Neutronic Analysis of Lead208-Bismuth Eutectic-Cooled Modified CANDLE Reactor with Core Geometry Variations

Lead208-Bismuth Eutectic (LBE)- cooled reactor is one of the Generation IV reactors. The fourth-generation has several objectives: economic competitiveness, inherent safety, minimize radioactive waste, and non-proliferation. Modified CANDLE burnup scheme is one solution to fulfil these objectives. The scheme does not need fuel enrichment and reprocessing Pu from LWR. Lead208-Bi eutectic was employed as a coolant of the reactor due to its low neutron absorption cross-section. However, However, it is necessary for surveying the neutronic core design. In this study, we compared the core neutronic among the variations of core geometry with equal volume of Pb208-Bi eutectic-cooled Modified CANDLE reactor. The calculation is performed by PIJ and CITATION of SRACs module. The Pancake cylindrical (larger diameter) has the highest K-eff and the lowest-peaking factor among of balance and tall cylindrical.

Effect of $$^{{\textbf{37}}}{\textbf{Cl}}$$ enrichment on neutrons in a molten chloride salt fast reactor

A molten chloride salt fast reactor (MCFR) is well suited to fuel breeding and the transmutation of transuranium (TRU) elements owing to its advantageous features of fast neutron spectrum and high TRU solubility. However, the neutron absorption cross section of $$^{35}{\hbox {Cl}}$$ is approximately 1000 times greater than for $$^{37}{\hbox {Cl}}$$, which has a significant impact on the neutron physical characteristics of a MCFR. Based on an automatic online refueling and reprocessing procedure, the influences of $$^{37}{\hbox {Cl}}$$ enrichment on neutron economy, breeding performance, and the production of harmful nuclides were analyzed. Results show that $$^{37}{\hbox {Cl}}$$ enrichment strongly influences the neutron properties of a MCFR. With natural chlorine, $$^{233}{\hbox {U}}$$ breeding cannot be achieved and the yields of S and $$^{36}{\hbox {Cl}}$$ are very high. Increasing the $${^{37}{\hbox {Cl}}}$$ enrichment to 97% brings a clear improvement in its neutronics property, making it almost equal to that corresponding to 100% enrichment. Moreover, when $${^{37}{\hbox {Cl}}}$$ is enriched to 99%, its neutronics parameters are almost the same as for 100% enrichment. Considering the enrichment cost and the neutron properties, a $$^{37}{\hbox {Cl}}$$ enrichment of 97% is recommended. Achieving an optimal neutronics performance requires 99% $$^{37}{\hbox {Cl}}$$ enrichment.

Effect of Corrosion of the Fuel Rod Construction Materials on the Radiation Safety of Nuclear Power Plants with PWR

The article considers the general concept of corrosion in accordance with GOST 5272–68 “Metal Corrosion”, the classification of the corrosion process, the stages of corrosion as energy function of the flow path of the corrosion process, the main indicators of the corrosion process. According to the forecasts of the International Monetary Fund and Focus Economics, the amount of funds that will be spent on counteracting corrosion and its consequences in selected industrialized countries has been estimated. The growth of funds invested in the counteracting the effects of metal corrosion in the Russian Federation for 2016–2019 is presented in the form of a diagram. The substantiation of the use of zirconium as a structural material for the shell of fuel rods has been fulfilled. The values of the thermal neutron absorption cross sections for various elements serving as structural elements for the core of a nuclear reactor are presented. Factors influencing the choice of alloying elements and their percentage in various alloys (Zr-2, Zr-4, ZIRLO™, M5®), which are the special development that reduce the corrosion rate, are also considered. The composition and mechanical properties of E110 and E635 alloys, which were used as materials for the fuel rods shell in the core of WWER-1200 reactors at the Belarusian NPP, are considered as well. The behavior of zirconium alloys E110 and E635 in the core is analyzed. The main factors that make a significant contribution to the corrosion process in actual operating conditions of zirconium alloys as fuel rods shell have been identified. The existing methods of preliminary special processing of fuel rods shells stored in the air for a long time before their receipt for assembly are presented. The structure of the oxide on the shells of alloys E110 and E635 oxidized in an autoclave is demonstrated.

Radiation shielding properties of silicon polymers

We have studied the X-ray and gamma radiation shielding parameters such as mass attenuation coefficient, linear attenuation coefficient, Half Value Layer (HVL), Tenth Value Layer (TVL), effective atomic number, electron density and specific gamma ray constant in some silicon polymers of different composition such as Polymer A- Poly dimethyl siloxane (C2H6OSi), Polymer B- Polymethyl Hydro silaxane (CH4SiO), Polymer C- Per hydro poly silaxane (H3SiN), Polymer D-poly dimethyl silaxane (C2H6Si), Polymer E− Methylsilses quioxane (C12H32O8Si8), and Polymer F- Silalkalyene polymer (SiC3H8). We have also measured the mass attenuation coefficient at gamma energies 84 keV to 1.3 MeV. We have also studied the neutron shielding properties such as coherent neutron scattering length, incoherent neutron scattering lengths, coherent neutron scattering cross section, incoherent neutron scattering cross sections, total neutron scattering cross section and neutron absorption cross sections in the silicon polymers. We have compared the shielding properties among the studied different silicon polymers. From the detail study, it is clear that the silicon polymer per hydro poly silaxane is good absorber for X-ray, gamma radiation and neutron. Hence, we suggest that the perhydropoly silaxane is good shielding material for X-ray, gamma and neutrons.

Neutronic Analysis to Replace Cobalt Radioisotope Facility by Iridium Radioisotope Facility at ETRR-2

The Egyptian Second Research Reactor (ETRR-2) is a pool-type reactor, 22 MW thermal, with 27 fuel elements loaded with 60Co production facility in the most relative highest flux position for the production of 200 Ci/g specific activity. The production of this specific activity needs a very long irradiation time and continuity of operation to produce useful quantities of 60Co over a reasonable period, which means that the reactor would have to operate 24 h a day, for 5 to 7 days a week. This requirement for the production of cobalt with the required specific activity is difficult to meet in ETRR-2, so this position needs to be reused for the production of other radioisotopes that require shorter irradiation times compared to cobalt. Iridium-192 is the most important radioactive isotope of iridium; it can be used in the production of “sealed sources” for industrial or medical applications. In this study, we did a full neutronic analysis of the ETRR-2 reactor core with iridium and with cobalt and compared both cases. We used two different models: a model using the MCNP code (Monte Carlo Neutron Photon), and another model using the WIMS/CITVAP code (a deterministic code). The models were validated with the results of the experiments done during the commissioning of the radioisotope production facility. We concluded that 500 g of iridium could be used instead of 577 g of cobalt in the core, and 24 molybdenum production plates would fulfil the fixed experiment design criteria, which is lower than 1200 pcm. The average axial/radial flux inside the tube was lower when using iridium disks than when using cobalt pellets because of the difference between the neutron absorption cross sections of 191Ir, 193Ir, and 59Co. When comparing the average radial flux inside the irradiation position near the edge of the iridium pellets inside the tube, we found that the flux would be higher for iridium than cobalt because of the empty part of the tube. We also calculated the power peaking factor over the whole core and found it was 2.12, which fulfilled the design criteria (must be less than 3).

Gamma, X-ray and neutron shielding properties of boron polymers

We have studied the X-ray and gamma radiation shielding parameters such as mass attenuation coefficient, linear attenuation coefficient, Half Value Layer (HVL), Tenth Value Layer (TVL), effective atomic number and electron density in some boron polymers of different boron based polymers [ Polymer A-PolyBorazylene (B 3 N 3 H 4 ), Polymer B- 4-Vinylphenyl Boronic acid (C 8 H 9 O 2 B), Polymer C- Borazine (B 3 N 3 H 6 ), Polymer D- 3-Acrylamidophenylboronic acid (C 9 H 10 BNO 3 ) Polymer E-Phenylethenylboronic acid (C 14 H 19 BO 2 ), Polymer F- 4-Aminophenylboronic acid (C 12 H 18 BNO 2 ) and Polymer G- 3- Aminophenylboronic acid (C 6 H 8 BNO 2 )]. We have also studied the neutron shielding properties such as coherent neutron scattering length, incoherent neutron scattering lengths, coherent neutron scattering cross section, incoherent neutron scattering cross sections, total neutron scattering cross section and neutron absorption cross sections in the boron polymers. We have compared the shielding properties among the studied different boron polymers. From the detail study, it is clear that the boron polymer Phenylethenylboronic acid is good absorber for X-ray, gamma radiation and neutron. Hence, we suggest that the boron polymer Phenylethenylboronic acid is good shielding material for X-ray, gamma and neutrons.

Directional reflection of cold neutrons using nanodiamond particles for compact neutron sources

Nanodiamond Particles (NDP) are new candidates for neutron reflection. They have a large scattering and low absorption cross-sections for low-energy neutrons. Very Cold Neutrons (VCN) are reflected from NDP with large scattering angles while Cold Neutrons (CN) have a quasi-specular reflection at small incident angles. A new scattering process has been added in Geant4 in order to examine the directional reflection of CN in an extraction beam made of NDP layer. Impurities in NDP are responsible for the up-scattered neutrons, especially hydrogen which has a large cross-section. Other impurities are also considered in Geant4 in order to produce a more accurate model for NDP scattering. The new scattering process is used to model a possible configuration of target-moderator-reflector in Compact Neutron Sources (CNS). A 13 MeV proton beam striking a beryllium target is chosen. Parahydrogen is placed as a cold moderator in order to produce CN. NDP are placed around the extraction beam for scattering the CN toward the exit of the beam. The results show that CN exiting the extraction beam can be increased thanks to the implemented NDP layer.

Corrosion analysis of post-heat treatment and post-weld SS316 with electrokinetic reactivation and cyclic polarization method

Austenitic stainless steel 316 has very high mechanical properties and corrosion resistance. This type of steel is widely used both in the nuclear and non-nuclear industries. In the nuclear industry, SS316 is used as a cladding material for uranium fuel due to its good corrosion and mechanical properties, and also low neutron absorption cross-section. In the Center for Nuclear Fuel Technology (PTBBN BATAN), it is used as material for the container of nuclear waste that is to be stored on Temporary Storage Installation of Spent Fuel (KHIPSB3 BATAN). SS316 is used as material for can to contain high-activity solid waste from the testing activity in Radiometallurgy Installation (IRM BATAN). The lid of the container is sealed with the GTAW welding process in order to tightly contain the solid waste. The main problem with the heat treatment and welding process of austenitic stainless steel is the occurrence of sensitization in a temperature range of 500-800°C. Therefore fully electrochemical analysis of SS316 stainless steel in various mediums has been conducted. SS 316 specimen was heat-treated to simulate the heat generated by the welding process. Sensitization analysis was conducted with qualitative and quantitative methods by EIS and EPR, and pitting corrosion resistance was by cyclic polarization method. The solution used for EIS and cyclic polarization test was 0 − 3.5% concentration of NaCl, and for EPR test was a mixture of H2SO4 and KSCN. Material characterization before and after corrosion testing was microstructure examination. The result of the corrosion test showed that heat treatment on the temperature range of 500-800°C caused sensitization. The corrosion test curve result showed that a heat treatment temperature of 650°C for 1 hour had the highest activation current. The lowest Rp value for SS316 specimen post heat treatment in 675°C was 69.410 ohm. The welded SS316 specimen had a higher corrosion current than that of the unwelded specimen. The microstructure of the welded specimen showed that there was intergranular corrosion particularly in the HAZ region. The effect of NaCl concentration on the cyclic polarization test showed that the higher the NaCl concentration, the more easily the pitting corrosion. The indication of pitting corrosion occurrence was evaluated by considering the Epit and Erp values. The lower Epit value meant that pitting corrosion was more easily to occur.

Effect of Cr content on microstructure and properties of Mo0.5VNbTiCrx high-entropy alloys

A combination of good mechanical properties, excellent corrosion resistance and superior irradiation resistance made high-entropy alloys (HEAs) as promising nuclear materials. In this work, the Mo0.5VNbTiCrx (x = 0–2.0) HEAs containing elements with low thermal neutron absorption cross-sections were designed and prepared. The alloys were produced by vacuum arc melting, followed by hot isostatic pressing at 1200 °C and 150 MPa for 2 h and annealing at 1200 °C for 72 h. The crystal structure, microstructure, hardness, and mechanical properties of the annealed alloys were investigated. The corrosion resistance in superheated steam was also preliminarily studied. When 0 ≤ x ≤ 0.75, the alloys consist of one body-centered cubic (BCC) phase. When 1.0 ≤ x ≤ 2.0, the alloys are composed of one BCC phase, one C15 Laves phase, and another face-centered cubic (FCC) phase. The effect of Cr content on the phase formation is discussed. The hardness and yield strength increase substantially with increasing Cr content, which could be attributed to the solid-solution-like strengthening. Moreover, the addition of Cr improves the anti-corrosion properties in superheated steam and the studied alloys exhibit better corrosion resistance than a currently used Zr-4 alloy. The Mo0.5VNbTi and Mo0.5VNbTiCr0.25 alloys exhibit the promising potential to be used as nuclear materials, as they show a good combination of mechanical properties and corrosion resistance.

Influence of 7Li enrichment on Th-U fuel breeding performance for molten salt reactors under different neutron spectra

To obtain an outstanding neutron economy and Th-U (Thorium-Uranium) breeding performance, a quite high 7Li enrichment up to 99.995% is generally required for the designs of molten salt reactors with FLiBe carrier salt. While such high 7Li enrichment would undoubtedly bring great technical challenges and high costs to the molten salt preparation. Meanwhile, the 6Li neutron absorption cross section differs greatly under different neutron spectra, which would probably mean the choice of 7Li enrichment can be different for the molten salt reactors with different neutron spectra. With the aim to provide an appropriate reference for the choice of 7Li enrichment in the engineering design of molten salt reactors, key parameters (including 233U inventory, neutron spectrum, temperature coefficient, 7Li enrichment, 7Li and 6Li neutron absorption rate, breeding ratio, relative 233U production and doubling time) and their time evolution are systematically explored based on the in-house developed code MSR-RS. The results show that 99.99% and 99.95% in 7Li enrichment introduce limited influence on Th-U breeding performance for a thermal and fast reactor respectively, and their temperature coefficient are negative. By trading off the cost and technologies required for enriching 7Li, the above two 7Li enrichment are therefore recommended.