Reactor Structural Materials Research Articles

The effect of residual mechanical stresses and vacancy defects on the diffusion expansion of the damaged layer during irradiation of BeO ceramics

The paper presents the results of a study on the application of Raman and UV spectroscopy methods to determine the structural damage kinetics in the near-surface layer of BeO ceramics caused by high-dose irradiation with He2+ ions. Interest in this type of ceramics is due to the combination of its structural and thermophysical parameters, making these ceramics one of the promising classes of materials for microelectronics and structural materials for nuclear reactors, with the possibility of operation in conditions of heightened radiation background. According to the conducted studies, it was established that with the irradiation fluence growth, changes in the nature of deformation structural distortions associated with the accumulation of residual mechanical stresses of tensile and compressive types are observed. At irradiation fluences of 1016-5×1016 Не2+/cm2, tensile stresses play a dominant role in structural distortions, while the value of compressive stresses at fluence growth makes up a small share in the overall nature of the deformations. Moreover, an elevation in the irradiation fluence above 5×1016 He2+/cm2 leads to a rise in the concentration of defects caused by the formation of oxygen vacancies, as well as He-VO type complexes, the presence of which is indicated by the halo intensity growth in the Raman spectra, as well as a change in the intensity of the absorption bands. Analysis of changes in thermophysical parameters revealed that a rise in structural distortions associated with the accumulation of complex defects results in thermal conductivity reduction and a deterioration in heat transfer processes associated with partial amorphization of the damaged layer. Moreover, the established direct relationship between the value of residual mechanical stresses and the degradation of thermal conductivity indicates the cumulative effect of destructive changes caused by irradiation, as well as the influence of diffusion mechanisms on the damaged layer thickness growth.

In-situ ETEM study of plasma-facing tungsten nanofuzz oxidation at atmospheric pressure: Microstructure evolution and substrate-free oxidation kinetics

To enable sustainable carbon-free fusion energy, managing reactor structural material degradation during normal operation as well as accident scenarios is vital. Tungsten (W) plasma-facing materials (PFMs) are susceptible to aggressive high-temperature oxidation during air-ingress fusion reactor accidents, yet there's a lack of oxidation kinetic data for irradiated tungsten. Here, we utilize atmospheric environmental transmission electron microscopy (ETEM) to present the first kinetic data for substrate-free W nanofuzz oxidation at 400 °C and 500 °C in 1 bar dry air. Comparison with pristine bulk W during the early parabolic stage suggests an irradiation-decelerated oxidation for W nanofuzz. Our time-resolved in-situ characterization reveals a durable amorphous surface oxide, likely promoted by high-flux He+ irradiation-induced surface defects, serving as an effective passivating layer that impedes nanofuzz oxidation onset. This surface oxide layer also interfaces well with newly formed orthorhombic WO3, facilitated by stress relief through He bubble shrinkage, providing lasting passivating protection throughout the nanofuzz parabolic oxidation. This new finding challenges conventional notions of irradiation's negative impact on metal oxidation, and calls for advanced characterization to enhance our understanding of fusion energy materials degradation, informed by further accident modeling.

Electrochemical measurement of the corrosion rates of structural materials in molten NaCl–MgCl2 at 580℃

Molten chlorides find varied applications in engineering disciplines, notably in the nuclear energy sector. However, the utilization of molten chlorides in molten salt reactors (MSRs) requires investigation of the corrosion behavior of the reactor's structural materials in such environments. In this study, we proposed a facile method for determining the corrosion rates of structural materials immersed in a high temperature NaCl–MgCl2 molten salt, based on electrochemical measurements by cyclic voltammetry. The method involves measuring the electrochemical signals of the soluble corrosion products as a function of time. EuCl3-induced corrosion of Inconel 625 and stainless steel 316 (SS 316) in molten NaCl–MgCl2 was monitored by electrochemical detection of Ni2+ and Fe2+. The corrosion rates were linearly dependent on the EuCl3 concentration, and first-order rate constants of 1.03 × 10-4 and 3.20 × 10-5 s−1 at 580 ℃ were obtained for Inconel 625 and for SS 316, respectively. Arrhenius plot analysis of the temperature dependence of the rate constant provided an activation energy of 43.3 kJ mol−1 for Inconel 625 corrosion in molten NaCl–MgCl2–EuCl3. The rate constant became temperature-invariant above 620 ℃ indicating the mass-transfer limitation of corrosion at higher melt temperatures. The proposed approach enables accurate and real-time evaluation of the corrosion rate in the molten salt, and accordingly, offers a straightforward solution for in-situ corrosion monitoring in MSR systems.

Interaction of CsOH with structural steel 316L and simulated products of the steel-cladding reaction under assumed conditions relevant for the accident at the Fukushima Dai-Ichi Unit 2 and 3

A high radiation dose rate under the shield plugs of the Units 2 and 3 reactors at the Fukushima Dai-Ichi Nuclear Power Station (1F), raised an important question on volatile radioactive Cs migration, adsorption and retention on various structural materials during the accident. The high dose rate interferes with decommissioning work. The purpose of this study is to present basic knowledge on contamination mechanisms of a BWR reactor structural materials and post-accident debris by Cs compounds to help decommissioning companies consider the ways of decontamination and debris handling during 1F debris retrieval. Having this goal, a series of tests on CsOH contamination mechanism under assumed accident conditions for Units 2 and 3 was performed. The results may improve the accuracy of estimating the situation inside the reactor from the perspectives of accident analysis and materials’ properties. The test temperature and compounds were chosen based on the probable accident scenario. In particular, Cs adsorption was checked using preoxidized 316L stainless steel, a product of control blade degradation (steel-B4C eutectic) and a product of steel-channel box and cladding interaction (Zr-Fe based multiphase material). Penetration ability of Cs into the oxidized and metallic layers of materials was evaluated. CsOH most probable adsorption temperature range and new Cs containing phases’ formation temperature range were followed.

The temperature effect on the formation and growth of helium-vacancy clusters in bcc iron: A comprehensive molecular dynamics simulation

Bubble generation can significantly degrade the performance of fusion reactor structural materials. In this work, the temperature effects on the initial clustering of helium (He) atoms, the formation of helium-vacancy clusters (HenVm) and the subsequent growth of the clusters into bubbles in bcc iron under different conditions were simulated by molecular dynamics. The results show that He behaviors are primarily dominated by coalescence at temperature range from 400 K to 500 K, which leads to a preference for He to bind together. Conversely, when temperature is higher than 500 K, dissociation become more prominent, causing large He clusters to be hard to form. Under the conditions of increasing He concentration and irradiation dose, the higher the temperature, the higher the density of HenVm clusters at He concentration around 100 appm. However, the temperature dependence reversed at high He concentration. The detailed analysis of formation cases revealed the percentage of nucleation sites for He bubbles at defects during He segregation. In addition, there is no significant difference among the He/V ratio of HenVm clusters formed at vacancies, interstitial dislocation loops (IDLs) and interstitial sites during the cluster growth. However, the He/V ratio of He bubbles growing on edge dislocation (ED) is overall lower than those growing at other formation sites, since the growth of HenVm clusters on ED requires a lower pressure to kick out self-interstitial atoms (SIAs) or SIA clusters. The average He/V ratio of He bubbles in bulk decreases as temperature rises from 400 K to 800 K, while that of He bubbles growing on ED is independent of temperature.

Body-centered cubic phase stability in cobalt-free refractory high-entropy alloys

Cobalt-free refractory high-entropy alloys (RHEAs) are strong contenders for structural materials in nuclear reactors because they do not exhibit cobalt activity under irradiation. The mechanical properties and thermal stability of RHEAs are primarily attributed to the solid solution phase, which is essentially the body-centered cubic (BCC) phase. The BCC phase formation rules thus became the basic criterion in the compositional design of RHEAs. In this paper, the BCC phase formation rules in cobalt-free RHEAs were determined via the calculation of six semiempirical parameters, namely, the entropy of mixing, enthalpy of mixing, atomic size difference, Ω-parameter, d-orbital energy level and valance electron concentration. The mixing enthalpy and atomic size differences are more effective than other semiempirical parameters for predicting BCC phase stability in cobalt-free RHEAs. The presence of aluminum is found to cause a notable alteration in the range of phase stability in cobalt-free RHEAs.

Impact of trace silicon on irradiation hardening and embrittlement of RAFM steel subjected to neutron irradiation

The content of silicon (Si) element in reduced activation ferritic/martensitic (RAFM) steels varies significantly among different candidates of fusion reactor structural materials. Si is not intentionally added to the RAFM steel but rather remains as a residue during the deoxidation process in steel fabrication. To control and reduce the Si content, removing processes and special treatments are required. The additional refining processes significantly increase the cost of steel fabrication. Currently, the role of residual or trace Si in neutron irradiation of steel remains unclear. In the presented study, two 9Cr-2 W RAFM steels with Si contents of 0.1 wt.% and 0.01 wt.% were fabricated and subjected to a neutron irradiation at 563 K with a neutron fluence of 4.8 × 1023m−2. The main results show that 0.1 wt.% Si can contribute to the suppression of irradiation hardening and embrittlement, as evidenced by the reduction in irradiation-increased hardness and yield strength, while maintaining preferable elongation. Microscopic analysis suggests that an appropriate content of Si helps refine the grains in RAFM steel, increases the ductile fracture region of the fracture surface, and effectively reduces the density of dislocation loops induced by neutron irradiation. The findings evidence that retaining a certain amount of trace Si should be an informed and rational choice in the design of RAFM steel.

Review on synergistic damage effect of irradiation and corrosion on reactor structural alloys

The synergistic damage effect of irradiation and corrosion of reactor structural materials has been a prominent research focus. This paper provides a comprehensive review of the synergistic effects on the third- and fourth-generation fission nuclear energy structural materials used in pressurized water reactors and molten salt reactors. The competitive mechanisms of multiple influencing factors, such as the irradiation dose, corrosion type, and environmental temperature, are summarized in this paper. Conceptual approaches are proposed to alleviate the synergistic damage caused by irradiation and corrosion, thereby promoting in-depth research in the future and solving this key challenge for the structural materials used in reactors.

REMOTE MONITORING OF RADIATION TEST TEMPERATURE REGIME AT A HIGH-CURRENT ELECTRON ACCELERATOR

Methods for evaluating the radiation resistance of structural materials of nuclear reactors have been developed and research is being carried out in NSC KIPT. As a part of these studies, a supercritical water convection loop with a chamber for irradiation of the tested samples by electron beam of the accelerator LU-10 (beam energy of 8…10 MeV, beam power up to 10 kW) was developed. A pyrometric system based on a DSLR camera (Digital Single-Lens Reflex Camera) has been created for the remote continuous non-intrusive control of the surface temperature of the irradiation cell in the range of 350…550°С. The work presents the results of calibration and use of the developed system in the conditions of a 500-hour run of sample irradiation in the loop.

Dissolution behavior and aging of iron–uranium oxide

Understanding the dissolution behavior of the fuel debris is necessary for the safe decommissioning of Fukushima Daiichi Nuclear Power Plants. The dissolution behavior of FeUO4 compounds formed by a high-temperature reaction of UO2 with iron, a stainless-steel component of reactor structural materials, was investigated under atmospheric conditions. The compounds were prepared in an electric furnace using U3O8 and Fe3O4 as starting materials, and their solid states were analyzed using X-ray diffraction, scanning electron microscopy–energy-dispersive X-ray spectroscopy, and X-ray absorption fine-structure spectroscopy. The fission products were produced via thermal-neutron irradiation. The concentration of nuclides dissolved in water was examined by performing static leaching tests of FeUO4 compounds for up to three months. A redox reaction was proposed to occur between trivalent Fe and pentavalent U ions in the early stage of FeUO4 dissolution. It was thermodynamically deduced that the reduced divalent Fe ion was finally oxidized into a trivalent ion in the presence of dissolved oxygen, and iron hydroxide limited the solubility of Fe. Meanwhile, the concentration of hexavalent U (i.e., uranyl ion) was limited owing to the presence of secondary minerals such as metaschoepite and sodium uranate and subsequently decreased, possibly owing to sorption on Fe oxides, for example. The concentrations of multivalent ions of fission products, such as Ru and Ce, also decreased, likely for the reason above. By contrast, the concentration of soluble Cs ions did not decrease. The validity of this interpretation was supported by comparing the results with the dissolution behavior of a reference sample (Fe-free U3O8).

Shrinkage and coalescence behavior of helium bubbles in Fe9Cr1.5W0.4Si F/M steel under the coupling effect of compression stress and temperature

As important irradiation defects, helium bubbles will threaten the stability and reliability of reactor structural materials. Extensive research has been conducted on the dynamic response of helium bubbles under thermal annealing, irradiation, or tension, but the behavior and mechanisms of helium bubble evolution under compression remain to be understood. Here, by adopting in-situ cyclic nanocompression testing on Fe9Cr1.5W0.4Si F/M steel containing helium bubbles at 673 K, we present direct experimental evidence for the shrinkage and coalescence behavior of helium bubbles under the coupling effect of compression stress and temperature for the first time. Unlike the stress-free condition where the bubble size shows an increasing trend, compressive stress causes the bubble to shrink. Moreover, the large-sized bubbles show a greater change in size than small-sized bubbles. At 673 K, there is no thermal motion of the bubbles in F/M steel, and it is the deformation and mutual squeezing of bubbles under compression that causes the coalescence of tightly arranged bubbles. Furthermore, helium bubbles can act as dislocation sources to emit dislocations under compression, and the emitted dislocations can in turn envelope around the helium bubbles, thus hindering bubble coalescence and leading to the formation of deformed bubbles.

Electronic effects on radiation damage in α-iron: A molecular dynamics study

Iron (Fe)-based alloys, which have been widely used as structural materials in nuclear reactors, can significantly change their microstructure properties and macroscopic properties under high flux neutron irradiation during operation, thus, the problems associated with the safe operation of nuclear reactors have been put forward naturally. In this work, a molecular dynamics simulation approach combined with electronic effects is developed for investigating the primary radiation damage process in α-Fe. Specifically, the influence of electronic effects on the collision cascade in Fe is systematically evaluated based on two commonly used interatomic potentials for Fe. The simulation results reveal that both electronic stopping (ES) and electron–phonon coupling (EPC) can contribute to the decrease of the number of defects in the thermal spike phase. The application of ES reduces the number of residual defects after the cascade evolution, whereas EPC has a reverse effect. The introduction of electronic effects promotes the formation of the dispersive subcascade: ES significantly changes the geometry of the damaged region in the thermal spike phase, whereas EPC mainly reduces the extent of the damaged region. Furthermore, the incorporation of electronic effects effectively mitigates discrepancies in simulation outcomes when using different interatomic potentials.

Development of machine learning and empirical interatomic potentials for the binary Zr-Sn system

Zirconium alloys are pivotal structural materials in nuclear reactors. Enhancing their properties and performance necessitates a profound understanding of the interactions between alloying elements and lattice defects. However, the atomic-level mechanisms that account for the effects of tin (Sn) on irradiation-induced defect evolution in zirconium (Zr) are not well understood yet. To bridge this gap, we conducted extensive first-principles calculations and utilized the obtained data to develop two interatomic potentials for the Zr-Sn system: a Zr-Sn moment tensor machine learning interatomic potential (MTP) and a modified embedded atom potential (MEAM) built upon existing unary ones. The reliability of the constructed potentials was validated through systematically comparing relevant physical properties of pure Zr, Sn metals, and Zr-Sn solid solutions, calculated using these potentials, to values obtained from density functional theory calculations and/or experiments. The developed potentials exhibit high fidelity and accuracy in capturing the behavior of the Zr-Sn system. Their applicability in terms of both reliability and computational efficiency was also discussed. These potentials provide a solid foundation for further exploration of the atomic-scale behavior of the Zr-Sn alloy systems, thereby facilitating the continuous optimization and application of the zirconium alloys.

Thermal Properties Investigation of FeCr Alloy Using Lammps Simulation: A Preliminary Study

At present, nuclear reactor technology that is widely used because of its proven reliability is the gen-III + nuclear reactor. Even if it is seen from the aspect of safety and reliability of this generation reactor, it has been proven, but because nuclear energy plays a vital role to meet the growing world energy needs, it is necessary to have a type of nuclear reactor that is tailored to those needs.The next generation of nuclear reactors must meet the requirements of fulfilling safety requirements, be flexible, a longer operating life (more than 60 years), more economical. In order for a reactor to produce higher power, a longer operating life and more economical, reactor structure materials which are capable of being operated at high temperatures are needed. The types of materials that are expected to meet these requirements include various types of ferritic / martensite steel, austenite, alloy steel containing nickel, and metal glass materials and ceramic materials. FeCr metal alloys are alloys that form the metals mentioned above, so it is important to conduct research both in simulation and experiment. Molecular Dynamics simulation of FeCr alloys using Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS) has been done to explore their thermodynamic characteristics such as heat treatment, solubility of Cr, atomic radial distribution function (RDF). The results of the simulation are illustrated using Visual Molecular Dynamics (VMD) code.
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Effect of He-ions irradiation on the microstructure and mechanical properties of TiTaNbZr refractory medium-entropy alloy film

In this work, TiTaNbZr refractory medium-entropy alloy (RMEA) films were irradiated by 60 keV helium (He) ions at the fluence of 5 × 1016 cm−2 - 2 × 1017 cm−2. After irradiation, a structural transformation of amorphous-BCC single phase, namely self-healing behavior, in TiTaNbZr RMEA films was revealed on account of the enhanced diffusion and redistribution of atoms, which could also lead to a decrease of defects. In this process, increased grain boundaries, acting as effective defect sinks, have inhibited the He bubbles within crystals. The short-range migration of atoms did not destroy the uniformity of elements dramatically. In addition, the correlation between the structural transformation and mechanical properties evolution in the RMEA films was also further clarified. No degradation of mechanical properties even for irradiation fluence as high as 2 × 1017 cm−2 due to the increased grain boundaries and self-healing mechanism. These results all indicate that the TiTaNbZr RMEA have outstanding irradiation resistance and potential applications in reactor structural materials. Meanwhile, the microstructural transformation and performance evolution of the TiTaNbZr RMEA explored in this study could provide more understanding of irradiation effect and a new perspective for the development of the alloys with self-healing ability in refractory high/medium entropy alloys (RHEAs/MEAs).

The response of silicon carbide composites to He ion implantation and ramifications for use as a fusion reactor structural material

Silicon carbide fibre–reinforced silicon carbide matrix (SiCf/SiC) composites are desirable in structural nuclear applications due to their low density, high temperature strength and tolerance to energetic radiation. We have subjected two industrial grades of SiCf/SiC to high energy He²⁺ ion irradiation, up to 10,000 appm implanted He at 700 °C, to determine their suitability for blanket structural applications in nuclear fusion. Minor crystallographic evolution is observed following irradiation, with evidence of phase localised stress, intragranular strain, and lattice swelling. This is attributed to vacancy production and subsequent He bubble formation, both intragranular and at grain boundaries.During post-irradiation annealing up to 1300 °C, varying degrees of He bubble evolution are observed, with the Si phase showing the highest level of instability. The fibres appear stable, with no detectable radiation-induced defects. Meanwhile, bubbles in the fibre coating and matrix grains grow and agglomerate. Despite this, no delamination or microcracking is observed.

Mathematical Calculation of Material Reliability Using Surface Roughness Feature Based on Plasma Material Interaction Experiment Results

The choice of reactor structural material design must take into account the TOKAMAK fusion reactors' structural reliability. Due to their high levels of heat and energy, fusion reactions have significant deformation effects, which reduce the efficiency of energy production in reactors. Material selection, erosion and damage, heat and stress management, reliability analysis, maintenance, and inspection are crucial elements in determining how reliable fusion reactors are. The focus of this work is on material selection and reliability analysis based on these parameters. The most common wall materials used in fusion reactors are tungsten, beryllium, steel, or graphite. It is advised to utilize aluminum because harmful Beryllium dust limits the study of this element. For this purpose, a target of aluminum samples is established with a plasma of He ions created by glow discharge. The dependability of the samples is determined by calculating the Weibull Distribution and measuring the roughness of the sample surfaces following exposure.

Швидкість анігіляції позитронів у точкових дефектах реакторних матеріалів у модифікованій моделі Тао - Елдрупа

Theoretical concepts of the positron annihilation process in structural materials of nuclear reactors, taking into account the peculiarities of their electronic structure, have been developed. The Tao - Eldrup model, which allows to analytically calculate the lifetime of a positron in a spherically symmetric potential well, has been modified for the case of a potential well of finite height, in order to expand the limits of the model's application. The dependence of the positron lifetime on the height and width of the potential well, which occurs at the point defects, was determined. The results obtained within the framework of the modified model provide important information for the analysis of positron lifetime spectra in irradiated materials and data for the verification of quantitative calculations by the method of density functional theory.

The role of nickel in forming a structure providing increased service properties of reactor structural materials

Nickel is an essential alloying element in steels used as structural materials in the most common nuclear power reactors of the VVER type. The paper considers reviews the results of structural studies of traditional and advanced materials of the vessels and internals of VVER-type reactors with high nickel contents in their compositions. It is shown that an increased nickel content (up to 5 wt.%) in the steels of VVER pressure vessels contributes to the formation of a more dispersed structure with a smaller size of substructural elements and an increased density of dislocations, as well as a higher volume density of carbide phases. The revealed features of the structure of the reactor pressure vessel steel with high nickel content have the prerequisites for improving the strength and viscoplastic properties due to the increased number of barriers both for the dislocation motion and brittle crack propagation. Using the example of materials for VVER internals, it is shown that the nickel content increased in them up to 25 wt.% contributes to an increase in the volume density of radiation defects (dislocation loops of various types) and radiation-induced phase precipitates (G-phase). As nickel increases from 10 to 25 wt.%, there is a tendency to reduce swelling, which contributes to less shape change of the components of the reactor vessel internals. At the same time, in the steel with the highest nickel content, the highest nickel content was found in the near-boundary regions of the matrix, which contributes to greater austenite stability and a lower probability of the formation of an embrittling α-phase. The data obtained in the work on the effect of nickel alloying on the steel structural phase state and service characteristics were used in the development of new materials for the vessels and internals of advanced reactors.

Effect of surface machining on the environmentally-assisted cracking of Alloy 182 and 316L stainless steel in light water reactor environments: results of the collaborative project MEACTOS

Abstract The main objective of the EU-funded project mitigating environmentally-assisted cracking through optimisation of surface condition (MEACTOS) was to gain knowledge on the ability of different surface machining procedures to mitigate environmentally-assisted cracking (EAC) in typical light water reactor structural materials and environments. Surfaces of cold-worked (CW) type 316L austenitic stainless steel and nickel-based weld metal Alloy 182 flat tapered tensile specimens were machined using different processes. EAC initiation susceptibility of these specimens was evaluated using constant extension rate tensile (CERT) tests under simulated boiling water reactor (BWR) and pressurized water reactor (PWR) conditions and assessed using constant load experiments. More than a hundred tests were performed covering about 10 years of autoclave testing time. Only minor or no measurable improvements in EAC initiation susceptibility as a function of surface treatments (grinding or advanced machining) compared to the standard industrial face milling were demonstrated. In most cases, the stress thresholds for EAC initiation determined in constant load tests confirmed the trend obtained from CERT tests. This paper summarises the most important results and conclusions concerning the EAC initiation behaviour for the CW 316L and Alloy 182 under reducing PWR and oxidizing BWR conditions.