Materials For Nuclear Applications Research Articles

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.

Molecular dynamics exploration of helium bubble nucleation and growth mechanisms in Fe70Ni11Cr19 austenitic stainless steel.

The growth of helium bubbles impacts structural integrity of materials in nuclear applications. Understanding helium bubble nucleation and growth mechanisms is critical for improved material applications and aging predictions. Systematic molecular dynamics simulations have been performed to study helium bubble nucleation and growth mechanisms in Fe70Ni11Cr19 stainless steels. First, helium cluster diffusivities are calculated at a variety of helium cluster sizes and temperatures for systems with and without dislocations. Second, the process of diffusion of helium atoms to join existing helium bubbles is not deterministic and is hence studied using ensemble simulations for systems with and without vacancies, interstitials, and dislocations. We find that bubble nucleation depends on diffusion of not only single helium atoms, but also small helium clusters. Defects such as vacancies and dislocations can significantly impact the diffusion kinetics due to the trapping effects. Vacancies always increase the time for helium atoms to join existing bubbles due to the short-range trapping effect. This promotes bubble nucleation as opposed to bubble growth. Interestingly, dislocations can create a long-range trapping effect that reduces the time for helium atoms to join existing bubbles. This can promote bubble growth within a certain region near dislocations.

HELIUM AND HYDROGEN EFFECTS IN STRUCTURAL MATERIALS FOR NUCLEAR APPLICATIONS

Displacement cascades produce a variety of defects under reactor conditions, but of particular concern is the simultaneous production of helium (He) and hydrogen (H), which enhances the degradation of structural materials. The overall majority of performed studies on helium and hydrogen interactions with materials were based on ion beam irradiation, which served as a convenient tool for the simulation of neutrons exposure over a variety of temperature and dose regimes due to the ability to widely vary and control the irradiation parameters. Experimental investigations of the hydrogen-defect interaction performed by thermal desorption spectroscopy, and the parameters of this interaction obtained by numerical simulations based on diffusion-trapping codes are debated. In this review, we also summarize previous studies on grain boundaries and nanoprecipitate effects on hydrogen transport in metals, as well as the role of hydrogen in the corrosion and cracking of steels. We discuss here issues of helium bubbles formation and some of the evidence for the synergistic effects of hydrogen and helium in the presence of displacement damage, and their influence on irradiation hardening and swelling. Particular attention was devoted to the features of hydrogen interaction with noble-gas bubbles, which were considered on the basis of most recent published data.

Effect of gamma-rays on the thermal and mechanical properties of epoxy/tungsten boride composites

High performance polymer composites are essential materials in nuclear applications where many components can receive high ionizing rays doses such as gamma-rays. In this work, two composites named E51WB and E51WB2 based on epoxy (E51) and tungsten borides (WB and WB2) were manufactured at room temperature and then exposed to gammarays doses up to 2000 KGy. The goal of this study was to investigate and compare the impact of gamma-rays on the chemical structure, mechanical and thermal properties of E51WB and E51WB2 composites using FTIR, SEM, TGA and tensile tests. As the results, the char yield (Yc) for E51WB gradually increased from 40 to 46% at 0, 1000 and 1500 KGy, then slightly drop to 44% at 2000 KGy. The E51WB sample displayed the strain value of 2.3% at 0 KGy, and it increased to 4.7 and 5 % at 1000 and 1500 KGy, respectively, while at 2000 KGy it decreased to 3.5%. This may be due to phenomena such as hydrogen abstraction, disproportion, degradation or the formation of new bonds. The comparison of the TGA results of E51WB and E51WB2 at 1000 and 1500 KGy respectively, leads to the conclusion that E51WB2 is thermally more resistant. In addition, by comparing the tensile strength of these two samples at the same irradiation doses, it can be stated that the E51WB2 composite is stronger than its counterpart E51WB.

Precipitation-hardening stainless steels: Potential use radiation shielding materials

In this study, it was focused on the research of new materials with high attenuation efficiency that can be used as radiation shielding material due to the incapability of conventional materials. In accordance with this purpose, the radiation shielding capabilities of four different precipitation-hardening stainless steels (PH-SSs), which are symbolized 15-5PH, 15-7PH, 17-4PH and 17-7PH, were determined both experimentally and theoretically in a wide range of radiation energy. For experimental measurements, two different detectors (Si(Li) and Na(Tl)) and twenty-three different photon energies (from 22.1 to 1332.5 keV) emitted from seven different radionuclides were used in narrow-beam transmission geometry. Phy-X/PSD software were used for theoretical calculations. The theoretically calculated parameters were determined both in the photon energies emitted by the radionuclides used in the experiment and in the continuous energy range of 0.015–15 MeV. Furthermore, in order to make a remarkable assessment of the integration of PH-SSs to nuclear energy applications, the all results obtained have been compared with the corresponding values of Fe-based steel concretes (steel-scrap (SS) and steel-magnetite (SM)), which are widely used as conventional shield materials in nuclear power plants. Moreover, in order to choose the best material to be used in nuclear applications among the PH-SSs, the results were evaluated comparatively in terms of both the characteristics of the materials and the magnitude of the radiation shielding parameters. Consequently, it was concluded that the material with the best radiation shielding performance among the examined PH-SSs was 15-7PH and the radiation shielding performance ranking was generally in the form of 15-7PH ≥ 15-5PH > 17-4PH > 17-7PH > SM > SS. Due to both high attenuation efficiency and superior properties for radiation shielding, it was determined that the PH-SSs can be used as new shielding materials in nuclear applications.

Fatigue damage evaluation of stainless AISI 347 steel by advanced microstructure-sensitive NDT analysis

There is an increasing need to test and evaluate the integrity of ageing components and structures in nuclear engineering. Since materials for nuclear applications are subjected to thermal, corrosive, tensile, and fatigue loads during their service life, a deep understanding of microstructural degradation is desirable, as it can be gained by nondestructive testing (NDT) methods.In this study, an innovative fatigue testing setup and NDT methods are combined to develop new, efficient and reliable concepts to evaluate fatigue life and damage mechanisms. For this purpose, specimens of metastable austenitic stainless AISI 347 (X6CrNiNb18-10, 1.4550) steel, relevant for pipe components, are tested in a setup allowing in-situ measurement of temperature, micromagnetic, electric and electrochemical NDT parameters. Strain-controlled constant amplitude and strain-increase tests were performed and various physical measurements were applied for a fatigue life analysis by short-time evaluation procedures. In-situ NDT parameters are linked to the microstructural evolution of the metastable austenite during cyclic loading. Extensive microstructural analysis through methods like electron backscatter diffraction have been performed in initial condition to help understanding data development versus life time. It is shown that an advanced test setup based on micromagnetic and electrochemical measurements is suitable to indicate fatigue damage-related mechanisms in a metastable austenitic material. In the future, these methods and findings could enable mobile testing systems to evaluate components in service regarding their remaining fatigue life.

Enhanced recombination suppresses the void swelling in bcc multi-component alloys

Multi-component alloys, e.g., high-entropy alloys, exhibit excellent mechanical and radiation tolerance properties, making them potential candidate materials in nuclear applications. In this work, the void swelling in a single-phase body-centered cubic (bcc) reduced activation multi-component alloy FeCrV under high dose Au-ion irradiation was investigated. It is found that FeCrV has a significantly suppressed void swelling compared to α-Fe, with voids of smaller size distributed in a narrower region. To understand the difference between these two alloys in the defect evolution and its possible effect on the void swelling, the diffusion of point defects in α-Fe and FeCrV was investigated. Combined Density Functional Theory (DFT) and kinetic Monto Carlo (kMC) method based on the environment-dependent vacancy migration energies was used to simulate the vacancy diffusion. It is shown that vacancy diffusion is enhanced, which is because vacancies prefer to migrate through the easy migration channels which have lower values of the wide migration energy distribution in multi-component alloys. Ab initio molecular dynamics (AIMD) was employed to simulate the interstitial diffusion. The results show the interstitial diffusion is sluggish in FeCrV due to the atomic traps existing that slow down interstitials’ migration. The migration process of point defects in bcc multi-component alloy shows clear chemical environment dependence. The closer mobilities of vacancies and interstitials enhance the local recombination of point defects and suppress the void swelling.

A thermodynamics-based damage model for the non-linear mechanical behavior of SiC/SiC ceramic matrix composites in irradiation and thermal environments

A damage model is developed and validated with experimental data for the non-linear mechanical behavior of SiC/SiC composite materials in nuclear applications. Cyclic thermal and mechanical loading associated with neutron irradiation effects of these composites leads to wide-spread and progressive micro-cracking that leads to loss of thermal conductivity and further enhancement of thermo-mechanical damage. A physics-based model of wide-spread micro-cracking is developed within the thermodynamic framework of continuum damage mechanics. Evolution equations for damage parameters that describe the growth of continuum damage are developed, where the material variables are obtained from experiments. The model novelty is in coupling mechanical, thermal, and irradiation damage through a consistent thermodynamic framework, including loss of thermal conductivity due to the evolution of mechanically induced micro-cracks. A number of thermo-mechanical experiments were conducted to confirm model assumptions. The model is shown to be validated with out-of-pile experiments, and then implemented using commercial finite element code COMSOL to the fuel cladding problem with normal and high radiation dose cases.

Preliminary study on Fe[sbnd]Gd alloys as binary alloys and master alloys for potential spent nuclear fuel (SNF) application

One of the biggest challenges of nuclear industries is to fabricate safe, stable and efficient materials for the storage and transportation of spent nuclear fuels (SNFs), these materials are known as neutron shielding and absorbing materials. Recently, Gd has received much attention as a neutron absorbing material in SNF applications owing to its high neutron absorption capability. Herein, we propose novel FeGd alloys as binary alloys and master alloys for potential materials in nuclear applications. FeGd alloys with 20, 55 and 80 wt% Gd were selected based on the FeGd phase diagram. All fabricated FeGd alloys had higher hardnesses than pure Fe due to the formed FeGd phases, and the Fe-rich phase had a higher hardness than the Gd-rich phase. On the other hand, corrosion resistance of the FeGd alloys decreased as the Gd content increased. Satisfactory Gd-based intermetallic dispersion and modifying effects were obtained by casting Fe and stainless-steel alloy with Fe80Gd. The average intermetallic size in the stainless-steel alloy remarkably decreased with a decrease in the average distance between Gd-based intermetallics. FeGd alloys possess requisite hardness, thermal conductivity, and corrosion resistance as well as dispersion ability of intermetallics; hence, they show potential as promising candidates for SNF applications.

Investigation of the Protective Capacities of Precipitation-Hardening Stainless Steels in terms of Charged and un-Charged Particle Radiation

In this study, it has been focused on the investigation of particle-shielding performances of precipitation-hardening stainless steels (PH-SSs). In line with this focus, the stopping powers and ranges values of four different PH-SSs (15−5PH, 15−7PH, 17−4PH and 17−7PH) for energetic charged particles (proton, electron and alpha particles) were carried out in the kinetic energy range of 0.01-20 MeV. In addition, the fast neutron removal cross-section values of PH-SSs examined for neutrons were calculated at 4.5 MeV. In order to achieve a remarkable conclusion about the particle-absorbing capacities of the PH-SSs investigated, all calculations were also performed for some concretes (steel-magnetite, steel-scrap and ordinary concretes) used as shielding materials in nuclear applications. The results obtained were comparatively presented as a function of kinetic energy of particles. In addition, the results obtained were evaluated in terms of both types of particle and phase structures of materials examined. According to the results obtained, it was observed that the all investigated parameters are independent of phase structures of PH-SSs, that the all calculated parameters for PH-SSs examined are very close to each other, and that the particle-shielding performances of PH-SSs under examination are better than comparison concretes. As a result of the data obtained from this study, it was observed that PH-SSs can be used as an alternative material in areas where particle radiation safety is required because of their superior characteristic and shielding properties.

Damage mechanics modeling of the non-linear behavior of SiC/SiC ceramic matrix composite fuel cladding

A damage model is developed and validated with experimental data for the non-linear mechanical behavior of SiC/SiC composite materials in nuclear applications. The model extends a continuum damage mechanics formalism, where explicit property degradation and volumetric swelling as a result of neutron irradiation are additionally included. The model is shown to be validated with out-of-pile experiments, and then applied to off-normal and high-burnup operating conditions associated with loss of cooling and extended radiation exposure, respectively. Optimized operating conditions aimed at reducing the fuel temperature included variation of the thermal conductivity and initial fuel-cladding gap thermal resistance. The nonlinear response along with continuum micro-cracking damage predictions are obtained and compared to the behavior during normal operating conditions. Numerical implementation of the model is based on 3-D Finite Element solutions.

Energy loss of 100 keV hydrogen isotopes in materials for nuclear applications

This work studies the energy loss by different mechanisms in materials of nuclear interest (Al, Ti, Fe) when these are irradiated with an hydrogenic beam. The materials chosen were aluminum, titanium and iron, and the selected beam energy was of the order of 100 keV. The mechanisms considered for energy loss are electron excitation and phonon production, for both of them two quantities are defined: one to compare energy deposition by a single ion beam in different materials, the other to characterize several beams impinging a given material.

Physical-Mechanical Properties of γ-Irradiated SiC Ceramics for Radioactive Wastes Immobilization

The interest in silicon carbide (SiC-based) ceramics and composites as matrix material for nuclear waste immobilization is grown up. Long-term chemical durability and radiation resistance of SiC are important factors for radionuclides immobilization. Advantages of SiC-based ceramics as structural materials in nuclear applications are the high-temperature properties, high density and reduced neutron activation. The use of radiation resistant materials is a strong requirement for safe and environmentally beneficial energy system. The SiC ceramics stability under irradiation for temperatures up to 1273 K is also very important for nuclear power applications. The SiC matrices doped by additives of Cr, Si were fabricated using High Speed Hot Pressing Method. Additives content was in the range from 0.5 to 3 wt %. Microstructural characteristics of silicon carbide ceramics were analyzed by X-ray diffraction (XRD), scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDX), and infra-red spectroscopy (IR) methods. The results of microcracking under indentation conditions were revealed the lack of cracks in the SiC ceramics with Cr additives before and after irradiation process. In addition, it was demonstrated that samples of SiC with alloying additives Cr and Si possess high mechanical parameters under γ-irradiation process. The strength of ceramics increases with the uniform and fine-grained structure formation. The modification of phase composition and mechanical properties of the SiC ceramics with Cr and Si additives under γ-irradiation were analyzed for further development of radiation resistant and matrix materials for radioactive wastes immobilization.

IRRADIATION EFFECTS ON ZR-2.5NB IN POWER REACTORS

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, “High Power Channel-type Reactor”) reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr-2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge.

Fabrication and characterization of joined silicon carbide cylindrical components for nuclear applications

The use of silicon carbide (SiC) composites as structural materials in nuclear applications necessitates the development of a viable joining method. One critical application for nuclear-grade joining is the sealing of fuel within a cylindrical cladding. This paper demonstrates cylindrical joint feasibility using a low activation nuclear-grade joint material comprised entirely of β-SiC. While many papers have considered joining material, this paper takes into consideration the joint geometry and component form factor, as well as the material performance. Work focused specifically on characterizing the strength and permeability performance of joints between cylindrical SiC–SiC composites and monolithic SiC endplugs. The effects of environment and neutron irradiation were not evaluated in this study. Joint test specimens of different geometries were evaluated in their as-fabricated state, as well as after being subjected to thermal cycling and partial mechanical loading. A butted scarf geometry supplied the best combination of high strength and low permeability. A leak rate performance of 2×10−9mbarls−1 was maintained after thermal cycling and partial mechanical loading and sustained applied force of 3.4kN, or an apparent strength of 77MPa. This work shows that a cylindrical SiC–SiC composite tube sealed with a butted scarf endplug provides out-of-pile strength and permeability performance that meets light water reactor design requirements.

Irradiation Effects in Materials for Nuclear Applications

Irradiation Effects in Materials for Nuclear Applications

Relationship between Microstructure and Mechanical Properties in ODS Materials for Nuclear Application

Relationship between Microstructure and Mechanical Properties in ODS Materials for Nuclear Application

Prospects of chemical vapor grown silicon carbide thin films using halogen-free single sources in nuclear reactor applications: A review

Abstract

Silicon Carbide Composite Fabrication and Mechanical and Thermal Performance for Nuclear Reactor Applications

Chemical vapor infiltrated (CVI) silicon carbide fiber reinforced-silicon carbide matrix (SiC-SiC) composite is a relatively new material and has come under increased scrutiny as an attractive material for nuclear applications. In these materials, the interplay between fiber and matrix leads to enhanced fracture toughness. However, due to the inherent directionality of the fibers, under-standing the mechanical and thermal performance is complex and requires careful characterization. At General Atomics, a laboratory has been established to develop these materials for a range of applications, in particular for the Energy Multiplier Module (EM2), which is a gas-cooled fast reactor.In this paper, the program of work that is being undertaken to fabricate and characterize SiC matrix com-posites is discussed. For mechanical testing, specialized fixtures have been developed to hold the thin ceramic composite specimens. For thermal testing, the purge gas species and flow rate have been identified as important parameters. Initial results on planar samples have been performed on SiC-SiC samples and hot pressed SiC samples.

Kinetics and Formation Mechanisms of a Nanostructured Fe-Y<sub>2</sub>O<sub>3</sub> Powder Prepared by Reactive Ball Milling

Reactive high energy ball-milling has known a growing interest from both fundamental and applied point of view. We focus here on the specific system Fe-Y2O3 metal-oxide nanocomposite because of its potential application to the synthesis of oxide dispersion strengthened steels, which are promising materials for nuclear applications. YFe3 and Fe2O3 were ball-milled during different milling durations in the stoichiometric proportions defined by the chemical reaction 2 YFe3 + Fe2O3 → 8 Fe + Y2O3. The obtained milled powder was characterized by XRD, SEM, TEM and their thermal behaviour was investigated by DTA. Through those characterizations, a Mechanically induced Self-propagating Reaction (MSR) was observed and several steps in the ball-milling process were identified: mixing of reactants, chemical reaction propagation, amorphization and refinement of the microstructure. The role of the milling intensity was also examined.