Displacement Damage Rate Research Articles

General displacement function for displacement damage cross-section calculation

The estimation of displacement damage rate requires the prior knowledge of displacement damage cross-section and displacement function is one part in the displacement damage cross-section calculation. The NRT model, which is used to calculate the displacement function for many years, always overestimates the number of displaced atoms in material. Although the arc-dpa model proposed recently can give correct results for the materials irradiated by neutron, there is still no suitable displacement function that can be used to accurately obtain both neutron and charged particles displacement cross-section comparing with experimental data. In the present work, a general displacement function is proposed, which is composed by power function model, arc-dpa model and BCA model. The model parameters are obtained for aluminum, iron, copper and tungsten. The displacement damage cross-sections for the materials irradiated by neutron and charged particles calculated using the general displacement function agree well with available experimental data.

Calculation of displacement damage cross-section for charged particles at energies up to 100 GeV

Accurate prediction of displacement damage is essential for the safety of nuclear materials irradiated by neutrons, gammas and charged particles. Displacement damage cross-sections are used for the calculation of displacement damage rate. With the development of accelerator facilities and spallation neutron sources, displacement damage rate calculations in these facilities require the availability of displacement damage cross-sections for charged particles. In the present work, new arc-dpa model parameters were obtained for aluminum, iron, copper and tungsten. The displacement damage cross-sections for protons, deuterons, and alpha-particles at energies up to 100 GeV are calculated using the standard NRT and the arc-dpa model with improved parameters for these materials. Calculated displacement damage cross-sections for charged particles calculated using the arc-dpa model with new parameters agree well with available experimental data.

Investigation on microscopic invasion characteristics and retention mechanism of fracturing fluid in fractured porous media

Reservoir damage caused by guar gum fracturing fluid and slick water seriously affects the subsequent oil and gas production. However, the invasion characteristics and retention mechanisms of fracturing fluids in the fracture-matrix zone are still unclear. In this work, a microscopic model reflecting the characteristics of the fracture-matrix zone was designed. Based on the microfluidic experimental method, the process of fracturing fluid invasion, flowback and retention in the fracture-matrix zone was investigated visually and characterized quantitatively. The factors and mechanisms affecting fracturing fluid retention in the fracture-matrix zone were analyzed and clarified. The results indicated that in the invasion process, the frontal swept range of slick water was larger than that of the guar gum fracturing fluid, and the oil displacement efficiency and damage rate were lower than those of the guar gum fracturing fluid under the same invasion pressure. With the increase in invasion pressure, the damage rate of slick water increased from 61.09% to 82.77%, and that of the guar gum fracturing fluid decreased from 93.45% to 83.36%. Before subsequent oil production, the invaded fracturing fluid was mainly concentrated in the medium-high permeability area of the fracture-matrix zone. The main resistance of slick water was capillary force, while that of the guar fracturing fluid was mainly viscous resistance. The fracturing fluid retention was most serious in the low permeability region and the region near the end of the fracture. The experimental and numerical simulation results showed that increasing the production pressure difference could improve the velocity field distribution of the fracture-matrix zone, increase the flowback swept range and finally reduce the retention rate of the fracture fluid. The retention mechanisms of slick water in the fracture-matrix zone include emulsion retention and flow field retention, while those of the guar gum fracturing fluid include viscous retention and flow field retention. Emulsion retention is caused by capillary force and flow interception effect. Viscous retention is caused by the viscous resistance of polymer, while flow-field retention is caused by uneven distribution of flowback velocity.

High-dose ion irradiation damage in Fe28Ni28Mn26Cr18 characterised by TEM and depth-sensing nanoindentation

One of the key challenges for the development of high-performance fusion materials is to design materials capable of maintaining mechanical and structural integrity under the extreme levels of displacement damage, high temperature and transmutation rates. High-entropy alloys (HEAs) and other concentrated alloys have attracted attention with regards to their performance under fusion conditions. In recent years, a number of investigations of the irradiation responses of HEAs have peaked the community’s interest in them, such as the work of Kumar et al. (2016), who examined Fe27Ni28Mn27Cr18 at doses as high as 10 dpa. In this work, we study Fe28Ni28Mn26Cr18 concentrated multicomponent alloy with irradiation doses as high as 20 dpa. We find the presence of Cr rich bcc precipitates in both the un-irradiated and in the irradiated condition, and the presence of dislocation loops only in the irradiated state. We correlate the features found with irradiation hardening by the continuous stiffness method (CSM) depth-sensing nanoindentation technique and see that the change in the bulk hardness increases significantly at 20 dpa for temperatures 450 °C. These results indicate that the alloy is neither stable as a single phase after annealing at 900 °C, nor particularly resistant to irradiation hardening.

A Slow Positron Beam Study of Helium Clustering in Oxide Dispersion Strengthened Steels

The present work provides an innovative approach to the experimental study of the radiation tolerance of oxide dispersion strengthened (ODS) steel comparing to its non-ODS variant. Inhomogeneous radiation damage, introduced by helium irradiation, was probed in a depth profile manner by a variable energy positron beam. Although numerous studies reporting the use of slow positron beam (SPB) experiments on ion-implanted materials have been published in the literature, only a few of them address the continuous distribution of microstructural damage introduced by ion beam. In this paper, we use SPB to probe a wide range of radiation damage in a single sample of each material. By evaluating positron diffusion in different parts of the Bragg peak, we were able to track the evolution of radiation-induced damage from a high density of small defects such as helium-vacancy clusters to large bubbles grown by coalescence. Combining with the transmission electron microscopy results and slow positron Doppler broadening spectroscopy, we offer new insight into different stages of microstructural bubble evolution in steels exposed to harsh radiation conditions. The obtained results show that the higher density of vacancy-type defects in the oxide dispersion strengthened steel, which act as recombination centres, suppress the formation of large defect agglomerations. While the TEM provides an indication of this effect in high displacement damage and helium production rates, positron annihilation spectroscopy elucidates the origin of this behaviour, improving the understanding of the fundamental mechanism of helium embrittlement.

Study on the Influence of Closed‐Cell Aluminum Foam on the Impact Performance of Concrete Pier after Equal Replacement with Stainless Steel Reinforcement

In this study, the impact test of two groups of reinforced concrete piers protected by closed‐cell aluminum foam is carried out by using the ultrahigh drop hammer impact test system. The purpose of this study is to explore the impact resistance and protective performance of closed‐cell aluminum foam under the impact load on the concrete bridge pier after replacing the ordinary reinforcement with stainless steel reinforcement. The study results show that the impact force is related to the overall stiffness of the specimen, as well as to the failure mode. When the impact velocity is less than 1.42 m/s, the closed‐cell aluminum foam is in an elastic or yielding stage. The change rate of impact force (231 and 97.5, respectively), tip displacement (33.5 and 18, respectively), and ultrasonic damage rate of the concrete in the two groups of specimen is relatively small, while the change rate of the two groups of specimen remains approximately consistent. In addition, when the impact is greater than 1.42 m/s and the closed‐cell aluminum foam is in the densification stage, the change rate of the impact force (increase from 231 to 819 and from 97.5 to 984.5), the tip displacement (increase from 33.5 to 67 and from 18 to 62), and ultrasonic damage rate of concrete are larger, which results in an increase in the dynamic response of the structure.

The design of the DONES lithium target system

In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,xn) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion Work Package Early Neutron Source as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES will employ a high speed liquid lithium jet struck by a 125 mA, 40 MeV deuteron beam to generate the intense neutron flux used to irradiate the material samples up to the desired level of displacement damage (˜10 dpa/fpy for iron in 0.3 l) and He production rates (˜10-13 appm He/dpa).In order to rapidly achieve a sound and stable design, a new configuration of the DONES target system based on the so-called integral concept has been proposed as reference solution in place of the former baseline design that envisaged a target assembly endowed with a replaceable back-plate, being the latter solution not yet fully qualified and thus not readily implementable.Moreover, following the outcomes of detailed dedicated analyses taking into account different aspects, the design of several components (Quench Tank, Lithium inlet pipe, Vacuum Chamber, etc.) has been changed.In this paper, a brief description of the current design of the DONES Target System is presented including all the above-mentioned aspects, showing the capability of the system to fulfil the prescribed requirements.

Effects of displacement damage and helium production rates on the nucleation and growth of helium bubbles – Positron annihilation spectroscopy aspects

Fe-12 wt% Cr model alloy samples were implanted by 250 keV He2+ ions to three different fluencies (3 × 1017, 9 × 1017 and 1.5 × 1018 cm−2) at T < 100 °C. In a depth profile manner, the implantation impact according to defined peak profile was investigated using variable energy slow positrons, with the primary focus on the 2–13 dpa region. The obtained data were compared to published data on Optifer IX steel samples [1] irradiated in the frame of a two-years irradiation program of the Swiss Spallation Neutron Source. Bi-modal defect distribution represented by two defect components in positron lifetime spectrum reveals two distinct helium bubbles growth mechanisms. While at the lower helium production rate of the spallation environment, the bubbles grow primarily by migration and coalescence, at the high production rates of helium in the implanted samples, the results indicate this growth is driven by Ostwald ripening mechanism. A competitive growth process via emission of interstitial atoms (clusters) is discussed in terms of low-temperature He implantations.

Opportunities and limitations for ion beams in radiation effects studies: Bridging critical gaps between charged particle and neutron irradiations

Ion beams are indispensible for understanding radiation effects in materials, enabling access to conditions challenging to perform with neutrons. However, ion irradiations can produce potential artifacts from beam rastering, high displacement damage rate effective “temperature shifts”, subthreshold collisions, high ionization rates, and non-monoenergetic primary knock-on atom energies. The influence of near-surface denuded zones and implanted ion effects is analyzed, including diffusional broadening effects. At high ion irradiation energies, “swift heavy ion” effects can lead to enhanced defect production or recovery. Ion energies of 10–50MeV typically offer a good compromise for minimizing near-surface, implanted ion, and swift heavy ion effects.

The Studies of Irradiation Hardening of Stainless Steel Reactor Internals under Proton and Xenon Irradiation

Abstract Specimens of stainless steel reactor internals were irradiated with 240 keV protons and 6 MeV Xe ions at room temperature. Nanoindentation constant stiffness measurement tests were carried out to study the hardness variations. An irradiation hardening effect was observed in proton- and Xe-irradiated specimens and more irradiation damage causes a larger hardness increment. The Nix-Gao model was used to extract the bulk-equivalent hardness of irradiation-damaged region and critical indentation depth. A different hardening level under H and Xe irradiation was obtained and the discrepancies of displacement damage rate and ion species may be the probable reasons. It was observed that the hardness of Xe-irradiated specimens saturate at about 2 displacement/atom (dpa), whereas in the case of proton irradiation, the saturation hardness may be more than 7 dpa. This discrepancy may be due to the different damage distributions.

Defects production and mechanical properties of typical metal engineering materials under neutron irradiation

Maintaining the safety and reliability of nuclear engineering materials under a neutron irradiation environment is significant. Atomic-scale simulations are conducted to investigate the mechanism of irradiation-induced vacancy formation in CLAM, F82H and α-Fe with different neutron energies and objective laws of the effect of vacancy concentration on mechanical properties of α-Fe. Damage of these typical metal engineering materials caused by neutrons is mainly displacement damage, while the displacement damage rate and the non-ionizing effect of neutrons decrease with the increase of neutron energy. The elastic modulus, yield strength, and ultimate strength of α-Fe are in the order of magnitude of GPa. However, the elastic modulus is not constant but decreases with the increase of strain at the elastic deformation stage. The ultimate strength reaches its maximum value when vacancy concentration in α-Fe is 0.2%. On this basis, decreasing or increasing the number of vacancies reduces the ultimate strength.

Characterization of neutron induced damage effect in several types of metallic multilayer nanocomposites based on Monte Carlo simulation

Metallic multilayer nanocomposites are known to have excellent interface self-healing performance when it comes to repairing irradiation damages, thus showing promise as structural materials for advanced nuclear power systems. The present study investigated the neutron irradiation displacement damage rate, spectra of the primary knocked-on atoms (PKAs) produced in the cascade collision, and the H/He ratio in four kinds of metallic multilayer nanocomposites (Cu/Nb, Ag/V, Fe/W, and Ti/Ta) versus neutrons’ energy. Results suggest that the three neutron induced damage effects in all multilayer systems increased with the increasing of incident neutrons’ energy. For fission reactor environment (1MeV), multilayer’s displacement damage rate is 5–10×1022dpa/(n/cm2) and the mean PKAs energy is about 16keV, without any noteworthy H/He produced. Fe/W multilayer seems very suitable among these four systems. For fusion reactor environment (14MeV), the dominant damage effect varies in different multilayer systems. Fe/W multilayer has the lowest displacement damage under the same neutron flux but its gaseous transmutation production is the highest. Considering the displacement damage and transmutation, the irradiation resistance of Ag/V and Ti/Ta systems seems much greater than those of the other two.

The Effect of Irradiation on Ni-Containing Components in CANDU®Reactor Cores: A Review

For nickel-containing alloys irradiated in thermalized neutron fluxes, the formation and reaction of59Ni with thermal neutrons can lead to time-varying changes in displacement damage rate, expressed as displacements per atom per second (dpa.s-1), gas formation and nuclear heating. Ni-rich alloys are used in PWR and BWR reactors as spacers within fuel assemblies but also as tensioning springs for these same assemblies. Flux thimbles have also been made from Ni-rich alloys in the past but are gradually being replaced by thimbles made from other alloys containing substantially less Ni. In a CANDU®reactor, Ni-alloys are used as tensioning springs, fuel channel spacers (in the form of garter springs) and as cable sheathing and core wires in flux-detector assemblies. Prediction of the irradiation processes that affect the functionality of these CANDU internals, such as irradiation embrittlement and irradiation creep, especially under conditions of extended operation, necessitates a consideration of the effect of the transmutation of Ni.Over the past 10 years, there has been a considerable increase in the understanding of the effects of irradiation on the properties of components made from Ni-rich alloys. The nuclear processes and the effects of irradiation damage on the performance of components made from these alloys will be described.

Irradiation creep of candidate materials for advanced nuclear plants

In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8dpa at displacement damage rates of 2–8×10−6dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400MPa at temperatures of 573, 673 and 773K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

Irradiation creep and microstructural changes of ODS steels of different Cr-contents during helium implantation under stress

Irradiation creep and microstructural changes of two ferritic ODS steels with 12% and 14% Cr have been studied by homogeneously implantation with helium under uniaxial tensile stresses from 40 to 300MPa. The maximum dose was about 1.2dpa (5000appm-He) with displacement damage rates of 1×10−5dpa/s at a temperature of 300°C. Irradiation creep compliances were measured to be 4.0×10−6dpa−1MPa−1 and 10×10−6dpa−1MPa−1 for 12 and 14Cr ODS, respectively. Subsequently, microstructural evolution was studied in detail by TEM observations, showing dislocation loops and bubbles distributed homogenously in the matrix. Some bubbles were attached to ODS particles. Finally, the effects of Cr content on irradiation creep and microstructural changes are discussed, including earlier results of a 19Cr ODS and a PM2000 ferritic steel.

Displacement damage rate dependence of defect cluster formation in α-Fe during irradiation

Formation kinetics of defect clusters in pure iron during irradiation has been numerically investigated by reaction rate theory, with focusing on nucleation process of vacancy clusters (voids) and self-interstitial-atoms (SIA) clusters under a wide range of atomic displacement damage rate (dpa rate) and temperature conditions. In the rate theory model, the size dependence of thermal stability of a defect cluster is treated for a wide range of cluster size. The numerical analysis shows that the nucleation processes of voids and SIA-clusters are quite different from each other. As to the voids, the nucleation rate of voids depends much on temperature and dpa rate, and has the individual peak temperature for each dpa rate, during which the peak temperature increases with increasing dpa rate. This tendency for void nucleation is similar to that for void swelling observed in experiments. As to the SIA-clusters, the nucleation rate of SIA-clusters does not depend much on temperature and has no peak temperatures because of the relatively high thermal stability of an SIA-cluster, indicating that the conventional model (di-interstitial model) is applicable to describe the nucleation of SIA-clusters in a wide range of temperature.

Neutronic analysis for the IFMIF target and test cell using a new CAD-based geometry model

This work presents neutronic analyses to support the IFMIF target and test cell (TTC) design in the framework of the Broader Approach activities. A very detailed Monte Carlo geometry model of IFMIF TTC based on the modular TTC concept was prepared directly from a CAD model by using the McCad conversion software which has been developed at KIT. The Monte Carlo code McDeLicious, which is an enhancement to MCNP5, was utilized and nuclear heating, displacement damage and gas production rates in the TTC vessel wall were calculated. The calculation result shows that there are two prominent peaks; downstream of the test modules due to the high energy neutron contribution for gas productions and upstream due to neutron back-streaming along the beam ducts. The result suggests it is very important in the neutronic analysis to consider the detailed configuration of TTC and test modules. The dose rate distribution during operation has been assessed for the rooms adjacent to TTC across thick surrounding walls. The necessary thickness for the shielding walls has been examined. The result demonstrates the substantial improvement in the shielding capability for the top access cell with the present TTC design.

Damage parameters of structural materials in fusion environment compared to fission reactor irradiation

Abstract Calculations were performed to quantify the damage parameters in the leading candidate structural and plasma facing materials when used in magnetic and inertial confinement fusion systems and when irradiated in fission reactors. The structural materials considered are ferritic steel, austenitic steel, vanadium alloy and SiC/SiC composite. Plasma facing materials included beryllium, tungsten, and carbon fiber composites. Atomic displacement damage and gas production rates are greatly influenced by the neutron energy spectrum. For the same neutron wall loading, atomic displacement damage is slightly lower in inertial fusion systems than in magnetic fusion systems but gas production is about a factor of 2 lower. In addition, much lower gas production is obtained in samples irradiated in fission reactors. The results help guide irradiation experiments in fission reactors to properly simulate the damage environment in fusion systems and facilitate extrapolating to the expected material performance in fusion systems.

A Solution of Displacement Damage Evaluation on Silicon Chip Under Space Environment

提出了一种模拟空间环境下半导体材料晶格损伤速率的方法, 计算了375km轨道高度不同纬度空间位置的飞行器模型中硅片产生晶格损伤的速率. 本方法考虑了宇宙射线的质子、正负电子、中子和γ射线成分, 建立了简单的多层飞行器结构模型, 并通过模拟给出了芯片上氕、氘、氚、氦、正负电子、正负π介子、γ射线及中子的损伤结果. 相比给出材料中的非电离能量损失, 更进一步给出了晶格损伤的产生速率, 为半导体器件的可靠性评估提供了进一步的参考数据.

Damage production and accumulation in SiC structures in inertial and magnetic fusion systems

Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is ∼10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.