High-temperature Neutron Irradiation Research Articles

Deuterium permeation performance of V4Cr4Ti/RAFM steel joint with in-situ formed carbide interlayer during diffusion bonding

Vanadium alloy (V4Cr4Ti) is a structural material used in future fusion reactors owing to its inherent advantages of high-temperature strength and neutron irradiation resistance. Oxygen pickup and high hydrogen isotope permeability are the two major concerns for their application in various types of blankets. To solve these issues, a bimetallic joint was designed with a CLF-1 layer of reduced-activation ferritic/martensitic (RAFM) steel to resist corrosion and a CLF-1/V4Cr4Ti bonding interface to suppress hydrogen isotope permeation. Samples were prepared from joints bonded via solid-state diffusion under hot isostatic pressing (HIP). Deuterium (D) gas permeation tests were performed at various temperatures in the range 400–600 °C at 100 kPa. After deuterium permeation tests, the thermal desorption spectra (TDS) of the samples were analyzed. And the permeation behavior of deuterium atom across the joints was schematically discussed. The results showed that bonding with steel could be an effective solution for resisting hydrogen isotope permeation through the V4Cr4Ti alloy. Thermal desorption spectrum (TDS) analysis of the tested samples revealed a desorption peak temperature equivalent to that of TiC. The formation of a 1 µm-wide B2 ordered α'- [Fe, (V, Ti)] phase layer, together with a large number of dispersed carbides around the interface, played a significant role in the deuterium permeation resistance. In addition, microstructural changes in the steel base material after the 900 °C bonding process were beneficial for suppressing deuterium diffusion.

Analysis of Shift in Nil-Ductility Transition Reference Temperature for RPV Steels Due to Irradiation Embrittlement Using Probability Distributions and Gamma Process

Reactor pressure vessel (RPV) steels are highly susceptible to irradiation embrittlement due to prolonged exposure to high temperature, high pressure, and intense neutron irradiation. This leads to the shift in nil-ductility transition reference temperature—∆RTNDT. The change in ∆RTNDT follows a certain distribution pattern and is impacted by factors including chemical composition, neutron fluence, and irradiation temperature. Existing empirical procedures can estimate ∆RTNDT based on fitting extensive irradiation embrittlement data, but their reliability has not been thoroughly investigated. Probability statistical distributions and the Gamma stochastic process were performed to model material property degradation in RPV steels from a pressurized water reactor due to irradiation embrittlement, with the probability models considered being normal, Weibull, and lognormal distributions. Comparisons with existing empirical procedures showed that the Weibull distribution model and the Gamma stochastic model demonstrate good reliability in predicting ∆RTNDT for RPV steels. This provides a valuable reference for studying irradiation embrittlement in RPV materials.

Microstructural changes induced in advanced tungsten grades under high temperature neutron irradiation

Recent studies addressing the change of mechanical properties after high temperature irradiation of tungsten have reported substantial increase of the ductile to brittle transition temperature as well as increase of the hardness. In this work, we performed microstructural analysis by means of transmission electron microscopy (TEM) on various tungsten grades earlier exposed to high temperature neutron irradiation at 1200 °C up to 1 dpa. The latter corresponds to the expected irradiation temperature on the surface of a tungsten monoblock during the steady state operation in ITER. TEM was applied to five tungsten (W) materials including ITER specification pure W and precipitate strengthened W alloys. The analysis of TEM data coupled with its discussion led to a number of important conclusions regarding the damage induced at 1200 °C neutron irradiation such as: (i) weak effect of recrystallization on the accumulation of the lattice damage; (ii) promising results on the tolerance of damage accumulation in W-Y2O3 grade and (iii) internal oxidation of TiC particles which may contribute to the huge increase of the irradiation induced hardening.

Interactive effect of thermal aging and proton irradiation on microstructural evolution and hardening of δ-ferrite in 308L stainless steel weld metal

In the harsh service environment of high temperature and intense neutron irradiation in water-cooled nuclear reactors, the austenitic stainless steel weld overlay cladding on the inner surface of the reactor pressure vessel suffers from thermal aging and irradiation damage simultaneously, which can induce microstructural evolution and hardening of the material. Since it is quite difficult to achieve this simultaneous process out of the pile, two kinds of combined experiments, i.e., post-irradiation thermal aging and post-aging irradiation were performed on 308 L stainless steel weld metals in this work. The interactive effect of thermal aging and proton irradiation on microstructural evolution and hardening of δ-ferrite in 308 L weld metal was investigated by combining atom probe tomography, transmission electron microscopy and nanoindentation tests. The results revealed that thermal aging could eliminate the dislocation loops induced by irradiation and affect the phase transition process by accelerating spinodal decomposition and G-phase precipitation, thus enhancing hardening of irradiated δ-ferrite. For the effect of irradiation on the microstructure and hardening of thermally aged δ-ferrite, however, intensive collision cascades can intensify G-phase precipitation and dislocation loop formation but decrease spinodal decomposition, leading to a limited effect on hardening of thermally aged δ-ferrite. Furthermore, the interaction of thermal aging and irradiation can promote G-phase precipitation. Meanwhile, the interaction can cause δ-ferrite hardening, which is mainly influenced by spinodal decomposition, followed by G-phase and dislocation loops, where spinodal decomposition and G-phase cause hardening by inducing strain fields.

Simulation of neutron irradiation-induced recrystallization of tungsten

<sec>Tungsten is the candidate for divertor target material in future fusion reactors. The tungsten divertor target is expected to long serve in a harsh environment of high temperature and high-energy neutron irradiation. This can lead to neutron irradiation-induced recrystallization of tungsten, thereby increasing the possibility of intergranular brittle failure and compromising the safe operation of the divertor. Thus, clarifying the mechanism of neutron irradiation-induced tungsten recrystallization is important. However, the current model, which only considers the irradiation-enhanced effect on recrystallization driving force, underestimates the irradiation effect on recrystallization compared with the results observed in recent high-temperature neutron irradiation experiments in the HFIR reactor. It indicates that other irradiation effects can also influence the recrystallization process.</sec><sec>In this study, we introduce the irradiation-enhanced grain boundary migration factor (<i>R</i>) into the established irradiation-induced recrystallization kinetic model, on the assumption that the grain boundary migration velocity is proportional to the self-diffusion coefficient. The simulation results show that after considering both irradiation-enhanced recrystallization driving force and grain boundary migration effect, the calculated half-recrystallization time (<inline-formula><tex-math id="M4">\begin{document}${t}_{{X}\text{}=\text{}0.5}$\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="16-20230531_M4.png"/></alternatives></inline-formula>) at 850 ℃ from the model matches the one obtained in the neutron irradiation experiment in the HFIR reactor. This result indicates that the irradiation-enhanced grain boundary migration effect is one of the important factors affecting irradiation-induced recrystallization. In addition, the difference between irradiated and unirradiated <i>t</i><sub><i>X</i>=0.5</sub> decreases with temperature increasing. This phenomenon is due to the fact that as the temperature increases, the contribution of irradiation defects to the driving force for recrystallization decreases owing to the irradiation defect recombination. Moreover, the increase of thermal activation diffusion coefficient is more significant than the increase of the irradiation-enhanced diffusion coefficient. These findings suggest that the thermal activation effect eventually dominates the recrystallization process over the irradiation effect as temperature increases.</sec>

Seeing through nuclear fuel: Three-dimensional, nondestructive X-ray microscopy and volumetric analyses of neutron-irradiated TRISO-coated fuel kernels

The three-dimensional (3D) characterization of nuclear fuel with X-ray microscopy has historically proven difficult, due to uranium’s high attenuation of easily accessible X-rays, both in a laboratory setting and at a synchrotron user facility. However, this imaging modality provides nondestructive information that can be used to investigate morphological changes arising from external stimuli (e.g., neutron irradiation, high-temperature testing). Using an appropriate X-ray energy spectrum and an adequate X-ray filter, suitable transmissions through properly sized nuclear fuel specimens can be achieved. Here, we present the methods and results of using a commercially available, laboratory-based X-ray microscope (XRM) to examine the extent of 3D morphological changes of tristructural isotropic (TRISO)-coated fuel particles, specifically uranium oxide/uranium carbide fuel kernels, after high-temperature neutron irradiation.

Mechanical properties of neutron-irradiated single crystal tungsten W(100) studied by indentation and FEM modelling

The aim of this work is to investigate the effect of high-temperature neutron irradiation on the mechanical properties of single-crystal tungsten W(100) studied by mechanical testing and modelling. Neutron irradiation up to 0.12 displacements per atom was performed at 600, 800, 900 and 1200 °C in the Belgium material test reactor BR2 at SCK•CEN in Mol. The mechanical properties of the irradiated tungsten were assessed by impulse excitation and depth-sensing indentation measurements. The values of Young's modulus, shear modulus and Poisson ratio are not affected by the irradiation. The mean hardness of the irradiated tungsten increases by 48, 45, 46, and 34% after irradiation at 600, 800, 900 and 1200 °C, respectively. The results obtained by the instrumented hardness tests were used as input for the finite element method model, applied to deduce the contributions coming from the neutron irradiation defects to the plastic deformation upon the indentation process, which include the impediment of gliding dislocations by sessile defects and modification of surface morphology due to the inhibited strain hardening ability through the dislocation-defect interaction. The validation of the fitted constitutive laws was realized by the crystal plasticity finite element method (CP-FEM) model which was applied to simulate the indentation load-depth curves, with certain assumptions on the irradiation-induced microstructure, and to calculate the distribution of stress under the indenter to investigate the extension of the plastic zone in the process of the indentation.

Anisotropy in the hardness of single crystal tungsten before and after neutron irradiation

In this work, the hardness of single crystal tungsten (W) after low, medium and high temperature neutron irradiation, reaching up to 1200°C, is studied. Micro-hardness tests with Vickers pyramid indenter are performed on reference and neutron irradiated samples by varying the orientation of the indenter on the single crystal tungsten contact surface. Due to the availability of multiple slip systems, the anisotropy of the hardness is characterized to allow the construction of uncertainty bands as well as dependence of the absolute value of the hardness on the orientation of the indenter wedges and principal crystallographic axes of the crystal. The applied neutron fluence and flux is representative for the ITER reactor, where tungsten is selected as armour material for plasma facing components. The neutron irradiation is performed in the BR2 reactor with mixed neutron spectrum up to 1 dpa (end of life fluence for ITER) but with appropriate measures to reduce the thermal neutron flux, such that the spurious transmutation into rhenium is reduced. The hardness study is accompanied with a detailed microscale examination to deduce the true indenter surface area. As a result, the true values of the hardness accounting for the pile-up effects are obtained as a function rotation angle of the indenter. The variation of the hardness as a function of indenter orientation before and after irradiation is discussed based on the indenter-slip system geometry. A simple but efficient methodology is proposed to deduce the true hardness and its angle-dependent variation for single crystal tungsten.

The effect of neutron irradiation in a helium atmosphere on the mechanical properties of 18Cr10NiTi austenitic steel

The microstructure and mechanical properties of 18Cr10NiTi austenitic steel were studied after high-temperature neutron irradiation of samples in a helium atmosphere. It was revealed during irradiation, helium penetrates into steel samples forming bubbles distributed on dislocations and along grain boundaries. Character of the temperature dependence of the mechanical characteristics of irradiated steel indicates a high-temperature radiation embrittlement of steel. The character of the mechanical properties temperature dependence change of irradiated steel points to its high-temperature radiation embrittlement.

D retention and depth profile behavior for single crystal tungsten with high temperature neutron irradiation

Single crystalline W (tungsten) samples irradiated at 633, 963 and 1073 K by neutrons to a damage level of 0.1 dpa were exposed to a high-flux D (deuterium) plasma at 673, 873 and 973 K, respectively, in TPE (Tritium Plasma Experiment) at INL (Idaho National Laboratory). Deuterium desorption was analyzed by TDS (Thermal Desorption Spectroscopy), and D depth profiles were determined by NRA (Nuclear Reaction Analysis) at SNL (Sandia National Laboratories). HIDT (Hydrogen Isotope Diffusion and Trapping) simulation code was applied to evaluate D behavior for neutron-damaged W at higher temperature.The D retention at depths up to 3 μm for the neutron-damaged sample at 673 K was two orders of magnitude larger than that for undamaged tungsten, and its D desorption spectrum had a single broad stage at around 900 K. As the neutron irradiation/plasma exposure temperature increased, D retention was largely reduced, and the desorption temperature was shifted to higher temperatures above 1100 K. The D depth profiles by NRA also showed D migration toward bulk by higher temperature irradiation, compared to undamaged W.The HIDT simulation indicated that the major binding energy of D was changed from 1.43 eV to 2.07 eV at higher neutron irradiation and plasma exposure temperatures, suggesting that some vacancies and small vacancy clusters would aggregate to form larger voids, or depopulation of weak traps at high D plasma exposure temperatures. It can be said that more stable trapping sites played dominant roles in the D retention at higher neutron irradiation and plasma exposure temperature. The binding energy by HIDT simulation was almost consistent with the reported value by TMAP, but the consideration of not only total D retention measured by TDS but also D depth profile by NRA led to the more accurate D behavior in neutron-damaged W.

Effect of high-temperature neutron irradiation on fracture toughness of ITER-specification tungsten

The effect of neutron irradiation on the fracture toughness of two commercially pure tungsten materials processed according to ITER specifications has been investigated for three doses: 0.08 dpa, 0.44 dpa, and 0.67 dpa at 600 °C. The choice of this temperature was motivated by its technological importance due to the risk of irradiation-induced embrittlement. The temperature of 600 °C is below the void swelling peak temperature (∼800 °C) and, at the same time, well above the ductile to brittle transition temperature (DBTT) of the reference material (∼300 °C). Neutron irradiation was performed in the BR2 material test reactor inside the fuel channel in order to limit the transmutation of rhenium and osmium close to the rates expected in a fusion environment. The results of the mechanical tests performed up to 600 °C show that the fracture toughness decreases with the increase in the irradiation dose for both tungsten products. The fracture surfaces of the non- and irradiated specimens were systematically analysed to determine the evolution of the failure mechanisms.

Equilibrium shapes and surface selection of nanostructures in 6H-SiC

The equilibrium shape of 6H-SiC nanostructures and their surfaces were studied by analyzing nano-void (∼10 nm) shapes, which were introduced in monocrystalline 6H-SiC by high-temperature neutron irradiation, using transmission electron microscopy. The nano-voids were determined to be irregular icosahedrons truncated with six {1¯100}, twelve {1¯103}, one smaller top-basal, and one larger bottom-basal planes, which suggests that {1¯100} and {1¯103} are the next stable surface class after the basal planes. The relatively frequent absence of the {1¯100} surface in the nano-voids indicated that the (1¯103¯) surface type is energetically rather stable. These non-basal surfaces were found not to be atomically flat due to the creation of nanofacets with half unit-cell height in the c-axis. The {1¯100} and {1¯103} surfaces were classified as two and four face types according to their possible nanofacets and surface termination, respectively. We also discuss the surface energy difference between the (1¯103¯) and (1¯103) face types in relation to the energy balance within the equilibrium, but irregular, polyhedron, in which the (1¯103) surface had double the surface energy of the (1¯103¯) surface (∼3900 erg/cm2).

Advanced materials for structural components of Indian sodium-cooled fast reactors

The performance of fast reactor (FR) working under high temperature, moderate pressure and intense neutron irradiation depends on its structural materials. Type 316 LN austenitic stainless steel with 0.02–0.03 wt.% carbon and 0.06–0.08 wt.% nitrogen is generally used for structural components of FR along with modified 9Cr–1Mo ferritic steel for steam generator. With an aim of enhancing the life of FR, efforts towards the enhancement of strength of the materials are been described. Nitrogen content in the 316 LN steel has been increased for optimizing the tensile, creep, fatigue and creep–fatigue interaction strengths. Both the tensile and creep strength were found to increase with nitrogen content; whereas the fatigue and creep–fatigue interaction strength have optimum values at around 0.12 wt.% of nitrogen. Based on the study, nitrogen content in 316 LN steel has been optimized at 0.11–0.13 wt.% for optimum combination of strengths. Weldability study has been performed and matching welding consumable has been developed.Creep strength enhancement of modified 9Cr–1Mo steel and its fusion welded joint has been addressed through microalloying the steel with boron along with the restriction on nitrogen content. The microalloying not only improves the creep strength of the steel also reduces the type IV cracking susceptibility of the fusion welded joint of the steel. The optimum contents of boron and nitrogen in the steel have been identified as around 60 ppm boron and 100 ppm nitrogen for better creep strength and type IV cracking resistance. The paper describes the challenges in developing the materials.

Propagation of High-Temperature Creep Cracks in Metals Subjected to Neutron Irradiation (A Survey)

We obtain a new equation for finding the rate of propagation of high-temperature creep cracks with regard for the influence of neutron irradiation. By using this equation, we construct a mathematical model for finding the residual life of a plate weakened by a macrocrack under the conditions of hightemperature creep and neutron irradiation. The application of the model is demonstrated for an analog of the Griffith problem. We establish a good correlation of the obtained analytic results with the available experimental data.

Neutron irradiation induced microstructural changes in NBG-18 and IG-110 nuclear graphites

This paper reports the neutron-irradiation-induced effects on the microstructure of NBG-18 and IG-110 nuclear graphites. The high-temperature neutron irradiation at two different irradiation conditions was carried out at the Advanced Test Reactor National User Facility at the Idaho National Laboratory. NBG-18 samples were irradiated to 1.54dpa and 6.78dpa at 430°C and 678°C respectively. IG-110 samples were irradiated to 1.91dpa and 6.70dpa at 451°C and 674°C respectively. Bright-field transmission electron microscopy imaging was used to study the changes in different microstructural components such as filler particles, microcracks, binder and quinoline-insoluble (QI) particles. Significant changes have been observed in samples irradiated to about 6.7dpa. The closing of pre-existing microcracks was observed in both the filler and the binder phases. The binder phase exhibited substantial densification with near complete elimination of the microcracks. The QI particles embedded in the binder phase exhibited a complete microstructural transformation from rosettes to highly crystalline solid spheres. The lattice images indicate the formation of edge dislocations as well as extended line defects bridging the adjacent basal planes. The positive climb of these dislocations has been identified as the main contributor to the irradiation-induced swelling of the graphite lattice.

Effects of manufacturing process on impact properties and microstructures of ODS steels

Oxide dispersion strengthened (ODS) steels are notable advanced alloys with durability to a high-temperature and high-dose neutron irradiation environment because of their good swelling resistance and mechanical properties under neutron irradiation. 9–12Cr-ODS martensite steels have been developed in the Japan Atomic Energy Agency as the primary candidate material for the fast reactor fuel cladding tubes. They would also be good candidates for the fusion reactor blanket material which is exposed to high-dose neutron irradiation. In this work, modification of the manufacturing process of 11Cr-ODS steel was carried out to improve its impact property. Two types of 11Cr-ODS steels were manufactured: pre-mix and full pre-alloy ODS steels. Miniature Charpy impact tests and metallurgical observations were carried out on these steels. The impact properties of full pre-alloy ODS steels were shown to be superior to those of pre-mix ODS steels. It was demonstrated that the full pre-alloy process noticeably improved the microstructure homogeneity (i.e. reduction of inclusions and pores).

Effect of Lead and Lead–Bismuth Eutectic Melts on the Fatigue Life of Steels of the Martensitic and Austenitic Classes

We study the effect of lead and lead-bismuth eutectic melts on the fatigue properties of martensitic Fe-13Cr and austenitic Fe-18Cr-10Ni steels within the temperature range 200-450°С. It is shown that the liquid-metal medium contributes to a significant decrease in the fatigue life of steels and the action of eutectic melt is more negative. The fatigue properties of austenitic Fe-18Cr-10Ni steel undergo weaker changes under the influence of lead and lead-bismuth eutectic melts. The realization of the projects of nuclear power plants of new generation requires the solution of a series of materials-science problems dealing with the development of reactor materials (1−4). It is supposed that steels of the austenitic and ferritic-martensitic classes may serve as the main structural materials of nuclear power plants. Despite the well-known susceptibility of steels of the austenitic class (based on the Fe−Cr−Ni system) to the de- velopment of vacancy porosity in the course of their high-temperature neutron irradiation, it is proposed to make the vessel of a BREST-type fast-neutron reactor and some in-vessel workpieces of chromium-nickel steel, which does not require additional thermal treatment after welding (5). Steels of the ferritic-martensitic class (based on the Fe−Cr system), due to their high thermal and mechanical characteristics, compatibility with the main cooling media, and lower sensitivity to swelling as compared with steels of the austenitic class, remain promising mate- rials for the first wall and blanket of fusion reactors, as well as for fuel elements, steam generators, and structural elements of the coolant pumps in the fast-neutron reactors (6, 7). The melts of heavy metals (Рb, Ві, and their eutectic mixture Pb−Bi), due to their nuclear and thermal prop- erties are regarded as promising cooling media for fast-neutron reactors (of the BREST-, SVBR-, and MBIR- types) and also for the subcritical hybrid systems controlled by accelerators (8−10). The problem of corrosion activity of these media in contact with steels is solved by using lead coolants with controlled oxygen contents, which promotes the formation of protective oxide films on the steel surface (11). Another hazard encountered in the course of operation of nuclear power plants is connected with liquid-metal embrittlement, which manifests itself in the degradation of the mechanical properties of structural materials in the course of their interaction with the melts of heavy metals. Unlike the studies of corrosion processes, the investigations of the influence of liq- uid-metal media on the degradation of the mechanical properties of steels are not systematic. This fact compli- cates the analysis of the mechanisms of crack initiation and propagation and the character of failures depending on the media and operating temperature and does not enable one to predict various manifestations of the interac- tions between liquid and solid metals and, hence, to predict the behavior of structural materials. Therefore, the analysis of the scientific aspects of the effect of lead melts (Рb, Рb−Bi) on the mechanical properties of steels of

The Stress and Reliability Analysis of HTR’s Graphite Component

The high temperature gas cooled reactor (HTR) is developing rapidly toward a modular, compact, and integral direction. As the main structure material, graphite plays a very important role in HTR engineering, and the reliability of graphite component has a close relationship with the integrity of reactor core. The graphite components are subjected to high temperature and fast neutron irradiation simultaneously during normal operation of the reactor. With the stress accumulation induced by high temperature and irradiation, the failure risk of graphite components increases constantly. Therefore it is necessary to study and simulate the mechanical behavior of graphite component under in-core working conditions and forecast the internal stress accumulation history and the variation of reliability. The work of this paper focuses on the mechanical analysis of pebble-bed type HTR's graphite brick. The analysis process is comprised of two procedures, stress analysis and reliability analysis. Three different creep models and two different reliability models are reviewed and taken into account in simulation. The stress and failure probability calculation results are obtained and discussed. The results gained with various models are highly consistent, and the discrepancies are acceptable.

Ag transport in high temperature neutron irradiated 3C–SiC

The effect of high temperature neutron irradiation on the ability of a Pd–Ag mixture to penetrate 3C–SiC has been investigated. Previous work has revealed enhanced Ag transport in SiC in the presence of the fission product Pd. In this work it has been shown that high temperature neutron irradiation leads to enhanced transport of this Pd–Ag mixture as compared to unirradiated material and that both high irradiation temperature and high neutron fluence is required for significant transport enhancement. The results suggest that grain boundary type and misorientation is not significantly altered by high temperature neutron irradiation, suggesting that these characteristics play only a secondary role in Pd–Ag transport.

Embrittlement of molybdenum–rhenium welds under low and high temperature neutron irradiation

The effect of low- and high-temperature neutron irradiation on the tensile strength, microhardness, and fracture mode has been studied for a series of Mo–Re welds with various Re concentrations. Radiation-induced hardening and concurrent ductility reduction are the key after-effects of neutron exposure. Low-temperature irradiation usually leads to a very hard embrittlement. The hardening effect is rather limited and unstable because of the lack of ductility. Irradiated specimens fail by brittle intergranular or transgranular fracture. The damaging effect of neutrons is less pronounced after high-temperature irradiation. The hardening of the matrix is rather high, but irradiated specimens still keep residual plasticity. High-temperature irradiation intensifies homogeneous nucleation of Re-rich phases, and this effect equalises the difference in mechanical properties between the different weld zones. A characteristic ductility loss exposure temperature was found to separate the temperature fields of absolutely brittle and relatively ductile behaviour. It usually varies between 850K and 1000K depending on the alloy composition and irradiation conditions.