Radiation Damage In Structural Materials Research Articles

Variable-energy cocktail beam technology for investigating synergistic damage in nuclear materials on LEAF platform

This study addresses the critical issue of synergistic radiation damage in structural materials of fusion reactors, focusing on the interaction between the displacement defects and transmutation-produced hydrogen and helium. These effects are simulated and investigated by employing the advanced multi-beam ion implantation capabilities of the LEAF (Low Energy high intensity highly charged ion Accelerator Facility) platform. Firstly, high-intensity cocktail beams, such as “4He+ and 56Fe14+" and “4He+ and 58Ni15+", are generated and characterized successfully. Then, a complex radiation environment is mimicked within the fusion reactors by applying variable-energy irradiation. Secondly, similar penetration depths for different ions, which are crucial for studying synergistic effects, are obtained by precisely controlling the energy of the cocktail beams through the innovative energy modulation system of the LEAF platform. Finally, the post-irradiation analyses, performed by using the transmission electron microscopy (TEM) and nanoindentation, revealed distinct microstructural changes and alterations in material properties, providing insights into the degradation mechanisms under irradiation. This work not only generates diverse and high-intensity “cocktail” ion beams but also achieves rapid energy switching of the beams. Further, the work is expected to pave the way for the implementation of a novel multi-beam irradiation technique in advanced heavy-ion linear accelerators, and also to provide innovative experimental methods and technical support for studying the synergistic effects of nuclear materials.

Data Driven In Situ TEM: A Path Towards Accurate Characterization of Radiation Damage in Structural Materials.

Data Driven In Situ TEM: A Path Towards Accurate Characterization of Radiation Damage in Structural Materials.

MECHANISMS OF RADIATION DAMAGE AND DEVELOPMENT OF STRUCTURAL MATERIALS FOR OPERATING AND ADVANCED NUCLEAR REACTORS

Safety of nuclear reactor (NR) and economic of nuclear power are determined to high degree by structural materials. Study of reasons of change of physical-mechanical properties of materials and of their dimensional stability under irradiation; determination of operation life of elements of nuclear power energetic assemblies in different conditions, selection and development of prospective materials with high radiation resistance are the main objectives of radiation material science. In the presented paper, mechanisms of radiation damage of structural materials for nuclear power and problems of development of radiation-resistant materials for operating and advanced NR of new generation are examined.

Carbon nanotube (CNT) metal composites exhibit greatly reduced radiation damage

Radiation damage of structural materials leads to mechanical property degradation, eventually inducing failure. Secondary-phase dispersoids or other defect sinks are often added to materials to boost their radiation resistance. We demonstrate that a metal composite made by adding 1D carbon nanotubes (CNTs) to aluminum (Al) exhibits superior radiation resistance. In situ ion irradiation with transmission electron microscopy (TEM) and atomistic simulations together reveal the mechanisms of rapid defect migration to CNTs, facilitating defect recombination and enhancing radiation tolerance. The origin of this effect is an evolving stress gradient in the Al matrix resulting from CNT transformation under irradiation, and the stability of resulting carbides. Extreme value statistics of large defect behavior in our simulations highlight the role of CNTs in reducing accumulated damage. This approach to controlling defect migration represents a promising opportunity to enhance the radiation resistance of nuclear materials without detrimental effects.

Memory access optimization of molecular dynamics simulation software Crystal-MD on Sunway Taihulight

The radiation damage effect of key structural materials is one of the main research subjects of the numerical reactor. From the perspective of experimental safety and feasibility, Molecular Dynamics (MD) in the materials field is an ideal method for simulating the radiation damage of structural materials. The Crystal-MD represents a massive parallel MD simulation software based on the key material characteristics of reactors. Compared with the Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS) and ITAP Molecular Dynamics (IMD) software, the Crystal-MD reduces the memory required for software operation to a certain extent, but it is very time-consuming. Moreover, the calculation results of the Crystal-MD have large deviations, and there are also some problems, such as memory limitation and frequent communication during its migration and optimization. In this paper, in order to solve the above problems, the memory access mode of the Crystal-MD software is studied. Based on the memory access mode, a memory access optimization strategy is proposed for a unique architecture of China's supercomputer Sunway TaihuLight. The proposed optimization strategy is verified by the experiments, and experimental results show that the running speed of the Crystal-MD is increased significantly by using the proposed optimization strategy.

Emulation of Radiation Damage of Structural Materials for Fission and Fusion Power Plants Using Heavy Ion Beams

The study is devoted to the methodology of simulation experiments for the analysis of radiation damage of structural materials of nuclear power plants by irradiation with heavy ions and subsequent analysis with use of the ultramicroscopy and nanoindentation methods. Details of the irradiation experiments in the TIPr accelerator (Institute for Theoretical and Experimental Physics) with ion energy of 101 keV/nucleon are given. Current approaches to the analysis of radiation-induced changes in the structural phase state of samples irradiated with ions with use of transmission electron microscopy and atom probe tomography are demonstrated. Models for the evaluation of radiation hardening based on microscopic changes, as well as the capabilities of the nanoindentation method for direct measurement of the hardening of a specimen layer irradiated by ions, are considered.

Application of arc-dpa model to estimate the primary radiation damage of structural materials by neutrons and the necessity of rescaling dpa versus final experimental damage correlations

The primary radiation damage (PRD) in structural materials is generally assessed in terms of the displacements per atom (dpa). The athermal recombination corrected displacements per atom (arc-dpa) model, used to predict PRD in materials has been incorporated in the Kalpakkam code CRaD. A recent parameterization of the arc-dpa model for various materials by Konobeyev et al. is available. This parameterization has facilitated a comparison between the predictions of the PRD levels in structural materials from NRT and arc-dpa models by using the CRaD code. A reference database of neutron dpa cross sections has been created by CRaD based on these two models by using the data from ENDF/B-VII.1 nuclear data library. The erstwhile design manuals of nuclear systems would have a specified limit of dpa towards life time integrity corresponding to methods and nuclear data used at that time. The new design manuals will have the same engineering damage correlated to the revised value of the dpa corresponding to the use of improved models of dpa and updated nuclear data for the same irradiation conditions. Therefore, the rescaling of dpa versus engineering damage is necessary, as the dpa is presently calculated using improved basic evaluated nuclear data and models for atom-displacement damage. The design life of nuclear systems is determined by the irradiation conditions, but the rescaling of dpa in the correlations of dpa versus damage makes sure that the updated dpa obtained with new data and methods is used correctly in design studies.

Emulation of reactor irradiation damage using ion beams

Progress in understanding radiation damage in structural materials is hampered by the lack of test reactors, long irradiations and high cost. Here we show that through strict control of experimental parameters and accounting for He production and damage-rate differences, the microstructure of ion-irradiated ferritic–martensitic steel closely resembles that created in-reactor across the full range of microstructure features. The level of agreement establishes for the first time the capability to tailor ion irradiation to emulate in-reactor radiation damage.

Calculation of displacement cross-sections for structural materials in accelerators using PHITS event generator and its applications to radiation damage

The displacement cross-sections, implemented in the particle and heavy ion transport code system (PHITS), have been calculated to estimate the radiation damage in structural materials used at accelerator facilities. The event generator in PHITS was used to calculate displacement cross-sections for 14 elemental targets and two practical alloys irradiated with neutrons at energies from 10−10 MeV to 3 GeV, and protons and deuterons at energies from several keV to 3 GeV. These displacement cross-sections were used to estimate the displacement per atom (DPA) in the target assembly at accelerator driven system (ADS) Target Test Facility (TEF-T) using 400-MeV protons, and in the Type 316 stainless steel target using 40-MeV deuterons. For the target assembly at TEF-T using 400-MeV protons, the DPA value for the proton component was approximately twice higher than that for the neutron component. For the Type 316 target using 40-MeV deuterons, radiation damage took place near the surface of Type 316 because of the Coulomb scattering effect between the incident deuterons and the material.

Solving the Puzzle of⟨100⟩Interstitial Loop Formation in bcc Iron

The interstitial loop is a unique signature of radiation damage in structural materials for nuclear and other advanced energy systems. Unlike other bcc metals, two types of interstitial loops, 1/2<111> and <100>, are formed in bcc iron and its alloys. However, the mechanism by which <100> interstitial dislocation loops are formed has remained undetermined since they were first observed more than fifty years ago. We describe our atomistic simulations that have provided the first direct observation of <100> loop formation. The process was initially observed using our self-evolving atomistic kinetic Monte Carlo method, and subsequently confirmed using molecular dynamics simulations. Formation of <100> loops involves a distinctly atomistic interaction between two 1/2<111> loops, and does not follow the conventional assumption of dislocation theory, which is Burgers vector conservation between the reactants and the product. The process observed is different from all previously proposed mechanisms. Thus, our observations might provide a direct link between experiments and simulations and new insights into defect formation that may provide a basis to increase the radiation resistance of these strategic materials.

The nature of high-energy radiation damage in iron

Understanding and predicting a material’s performance in response to high-energy radiation damage, as well as designing future materials to be used in intense radiation environments, requires knowledge of the structure, morphology and amount of radiation-induced structural changes. We report the results of molecular dynamics simulations of high-energy radiation damage in iron in the range 0.2–0.5 MeV. We analyze and quantify the nature of collision cascades both at the global and the local scale. We observe three distinct types of damage production and relaxation, including reversible deformation around the cascade due to elastic expansion, irreversible structural damage due to ballistic displacements and smaller reversible deformation due to the shock wave. We find that the structure of high-energy collision cascades becomes increasingly continuous as opposed to showing sub-cascade branching as reported previously. At the local length scale, we find large defect clusters and novel small vacancy and interstitial clusters. These features form the basis for physical models aimed at understanding the effects of high-energy radiation damage in structural materials.

Combining in situ transmission electron microscopy irradiation experiments with cluster dynamics modeling to study nanoscale defect agglomeration in structural metals

We present a combinatorial approach that integrates state-of-the-art transmission electron microscopy (TEM) in situ irradiation experiments and high-performance computing techniques to study irradiation defect dynamics in metals. Here, we have studied the evolution of visible defect clusters in nanometer-thick molybdenum foils under 1MeV krypton ion irradiation at 80°C through both cluster dynamics modeling and in situ TEM experiments. The experimental details are reported elsewhere; we focus here on the details of model construction and comparing the model with the experiments. The model incorporates continuous production of point defects and/or small clusters, and the accompanying interactions, which include clustering, recombination and loss to the surfaces that result from the diffusion of the mobile defects. To account for the strong surface effect in thin TEM foils, the model includes one-dimensional spatial dependence along the foil depth, and explicitly treats the surfaces as black sinks. The rich amount of data (cluster number density and size distribution at a variety of foil thickness, irradiation dose and dose rate) offered by the advanced in situ experiments has allowed close comparisons with computer modeling and permitted significant validation and optimization of the model in terms of both physical model construct (damage production mode, identities of mobile defects) and parameterization (diffusivities of mobile defects). The optimized model exhibits good qualitative and quantitative agreement with the in situ TEM experiments. The combinatorial approach is expected to bring a unique opportunity for the study of radiation damage in structural materials.

Studying Radiation Damage in Structural Materials by Using Ion Accelerators

Radiation damage in structural materials is of major concern and a limiting factor for a wide range of engineering and scientific applications, including nuclear power production, medical applications, or components for scientific radiation sources. The usefulness of these applications is largely limited by the damage a material can sustain in the extreme environments of radiation, temperature, stress, and fatigue, over long periods of time. Although a wide range of materials has been extensively studied in nuclear reactors and neutron spallation sources since the beginning of the nuclear age, ion beam irradiations using particle accelerators are a more cost-effective alternative to study radiation damage in materials in a rather short period of time, allowing researchers to gain fundamental insights into the damage processes and to estimate the property changes due to irradiation. However, the comparison of results gained from ion beam irradiation, large-scale neutron irradiation, and a variety of experimental setups is not straightforward, and several effects have to be taken into account. It is the intention of this article to introduce the reader to the basic phenomena taking place and to point out the differences between classic reactor irradiations and ion irradiations. It will also provide an assessment of how accelerator-based ion beam irradiation is used today to gain insight into the damage in structural materials for large-scale engineering applications.

Interface-enabled defect reduction in He ion irradiated metallic multilayers

Metallic multilayers are good model systems to explore the effects of heterophase interfaces in reducing radiation damage in structural materials. We summarize recent studies on radiation damage in immiscible face-centered cubic/body-centered cubic metallic multilayers, in particular Cu/V and Cu/Nb. These multilayers have shown unique characteristics compared to bulk metals under irradiation, including several orders of magnitude higher He solid solubility, dramatic reduction of bubble density, interface confined growth of He bubbles, and much lower radiation hardening. The mechanisms for interface enhanced radiation tolerance are briefly discussed.

Challenge of nuclear data for innovative nuclear energy

The article devoted to assessment of present-day demand to nuclear data for transmutation problem, including the discussion of required accuracies, status and perspectives of nuclear data evaluation and development of nuclear models. The effect of nuclear data uncertainties on radiation damage of structural materials is discussed. An analysis of ISTC projects related to nuclear data measurement and evaluation is presented. The recommendations for differential, integral experiments and recommendations on the evaluated data preparation are presented.

Structural factor and effects of low-temperature radiation damage in structural materials

The effect of the structural and phase state of steels and alloys based on iron and nickel on the recombination of vacancies and interstitial atoms in the process of neutron irradiation at 300–350°C is estimated.

Long-life Water Cooled Small Reactor with U—Np—Pu Fuel

The paper presents the advanced concept of a long-life small light water reactor in which the fuel irradiation time is comparable with reactor life-time. The equilibrium analysis reveals that the U-Np-Pu fuel with unique neutronic properties allows to keep sufficient criticality up to burnup value about 140GWd/tHM. The fuel recycle does not lead to additional Pu accumulation. Both Pu and Np are well protected against un-controlled proliferation by a large fraction of 238Pu in their mixture. To improve the reactor safety, the wider fuel pin lattice was applied. The radiation damage of structural materials is within the stainless steel limitation.

On the exploration of innovative concepts for fusion chamber technology

This study, called APEX, is exploring novel concepts for fusion chamber technology that can substantially improve the attractiveness of fusion energy systems. The emphasis of the study is on fundamental understanding and advancing the underlying engineering sciences, integration of the physics and engineering requirements, and enhancing innovation for the chamber technology components surrounding the plasma. The chamber technology goals in APEX include: (1) high power density capability with neutron wall load >10 MW/m 2 and surface heat flux >2 MW/m 2, (2) high power conversion efficiency (>40%), (3) high availability, and (4) simple technological and material constraints. Two classes of innovative concepts have emerged that offer great promise and deserve further research and development. The first class seeks to eliminate the solid “bare” first wall by flowing liquids facing the plasma. This liquid wall idea evolved during the APEX study into a number of concepts based on: (a) using liquid metals (Li or Sn–Li) or a molten salt (Flibe) as the working liquid, (b) utilizing electromagnetic, inertial and/or other types of forces to restrain the liquid against a backing wall and control the hydrodynamic flow configurations, and (c) employing a thin (∼2 cm) or thick (∼40 cm) liquid layer to remove the surface heat flux and attenuate the neutrons. These liquid wall concepts have some common features but also have widely different issues and merits. Some of the attractive features of liquid walls include the potential for: (1) high power density capability; (2) higher plasma β and stable physics regimes if liquid metals are used; (3) increased disruption survivability; (4) reduced volume of radioactive waste; (5) reduced radiation damage in structural materials; and (6) higher availability. Analyses show that not all of these potential advantages may be realized simultaneously in a single concept. However, the realization of only a subset of these advantages will result in remarkable progress toward attractive fusion energy systems. Of the many scientific and engineering issues for liquid walls, the most important are: (1) plasma–liquid interactions including both plasma–liquid surface and liquid wall–bulk plasma interactions; (2) hydrodynamic flow configuration control in complex geometries including penetrations; and (3) heat transfer at free surface and temperature control. The second class of concepts focuses on ideas for extending the capabilities, particularly the power density and operating temperature limits, of solid first walls. The most promising idea, called EVOLVE, is based on the use of a high-temperature refractory alloy (e.g. W–5% Re) with an innovative cooling scheme based on the use of the heat of vaporization of lithium. Calculations show that an evaporative system with Li at ∼1 200°C can remove the goal heat loads and result in a high power conversion efficiency. The vapor operating pressure is low, resulting in a very low operating stress in the structure. In addition, the lithium flow rate is about a factor of ten lower than that required for traditional self-cooled first wall/blanket concepts. Therefore, insulator coatings are not required. Key issues for EVOLVE include: (1) two-phase heat transfer and transport including MHD effects; (2) feasibility of fabricating entire blanket segments of W alloys; and (3) the effect of neutron irradiation on W.

Long-life Water Cooled Small Reactor with U-Np-Pu Fuel.

The paper presents the advanced concept of a long-life small light water reactor in which the fuel irradiation time is comparable with reactor life-time. The equilibrium analysis reveals that the U-Np-Pu fuel with unique neutronic properties allows to keep sufficient criticality up to burnup value about 140GWd/tHM. The fuel recycle does not lead to additional Pu accumulation. Both Pu and Np are well protected against un-controlled proliferation by a large fraction of 238Pu in their mixture. To improve the reactor safety, the wider fuel pin lattice was applied. The radiation damage of structural materials is within the stainless steel limitation.

Atomic processes during damage production and defect retention

Fundamental atomic processes during production and retention of radiation damage in structural materials are reviewed. Firmly established damage mechanisms are outlined, and some of the major uncertainties exposed. Damage production under neutron and ion irradiation proceeds through the elementary step of collision cascade formation, with further modification of the nascent cascade structure by inter-, and intra-cascade interaction events. Under fusion neutron irradiation, single cascade damage production occurs within 10 ps in most metals, with spatial extent of several tens to hundreds of nano-meters. Damage accumulation takes place over much longer time and length scales, during which multiple cascades interact leading to further re-structuring of the incipient cascade. Areas which require further investigation are: (1) the mechanisms of vacancy cluster collapse at the cascade core; (2) the athermal nucleation of interstitial loops at the periphery; (3) the migration of interstitial atoms and clusters emanating from the cascade; (4) the replacement of the dpa with a more physical measure of damage retention; (5) cascade structure for high PKA energies (above 20 keV); and (6) Inter- and Intra-cascade interactions, (7) and finally, cascade dynamics in damaged materials at high dose.