Irradiated Structural Materials Research Articles

Comparison of Monte Carlo Tools for Displacement Calculations for Protons and Heavy Ions Irradiation on Iron and EUROFER97

Abstract Ion irradiation of structural materials for the current and future generation nuclear reactors is a widespread technique used to simulate neutron irradiation in the context of material science research. During the designing phase of irradiation experiments Monte Carlo transport codes are often employed to quantify the radiation exposure, by estimating the displacements per atom (dpa) quantity. From the available displacement calculation codes, four popular choices were selected: cern-fluka, mcnp, phits, and srim. These codes will be used to simulate the irradiation effects of proton and Fe ions on pure iron (G379) and EUROFER97 alloy (a common material in fusion applications) targets. Results coming from the different software are presented in the form of damage profiles and of number of displacements as a function of the primary beam energy. This study identifies discrepancies between the codes' outputs and provides optimized simulation setup parameters.

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe3+ and 3.42 MeV energy degraded He2+) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. The comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

The role of stacking fault tetrahedra on void swelling in irradiated copper

A long-standing and critical issue in the field of irradiated structural materials is that void swelling is significantly higher in face-centered cubic-structured (fcc) materials (1% dpa−1) as compared to that of body-centered cubic-structured (bcc) materials (0.2% dpa−1). Despite extensive research in this area, the underlying mechanism of the difference in swelling resistance between these two types of materials is not yet fully understood. Here, by combining atomistic simulations and STEM imaging, we find stacking fault tetrahedra (SFTs) are the primary cause of the high swelling rate in pure fcc copper. We reveal that SFTs in fcc copper are not neutral sinks, different from the conventional knowledge. On the contrary, they are highly biased compared to other types of sinks because of the SFT-point defect interaction mechanism. SFTs show strong absorption of mobile self-interstitial atoms (SIAs) from the faces and vertices, and weak absorption of mobile vacancies from the edges. We compare the predicted swelling rates with experimental findings under varying conditions, demonstrating the distinct contributions of each type of sink. These findings will contribute to understanding the swelling of irradiated structural materials, which may facilitate the design of materials with high swelling resistance.

HPC atomic simulations of defect formation and kinetic Monte Carlo simulation of material ageing

Kinetic Monte Carlo (KMC) simulations are developed to simulate microstructure evolution under irradiation of structural materials of nuclear power plants. Methods based on rigid lattice atomic KMC and object KMC, despite some approximations, present the advantage to reach significant time, hence irradiation doses. The physical inputs such as the primary damage and defect cluster properties are the results of intensive atomic simulation on high performance computers. Here, atomic KMC is applied to model microstructure evolution of reactor pressure vessel steels and object KMC is applied to zirconium cladding materials.

Neutron availability in the Complementary Experiments Hall of the IFMIF-DONES facility

The IFMIF-DONES facility will be dedicated to the irradiation of structural materials planned for the use in future fusion reactors such as DEMO (Demonstration Fusion Power Plant). The potentialities of the IFMIF-DONES facility to complement its principal purpose by other experiments that would open the facility to other communities is addressed in this work. It concerns a study based on simulations to evaluate the neutronic performances of IFMIF-DONES in an hall dedicated to complementary experiments where neutrons can be transported. With the simple beam tube geometry of 4.5 cm entrance diameter studied in this work we have shown that a collimated fast-neutron beam of about 2 1010 n/cm2/s is available in the hall. Adding a moderator in the hall with neutron extraction lines would allow to get thermal neutron beams of about 106-107 n/cm2/s for dedicated experiments. The results show that IFMIF-DONES has the potentialities to be a medium-flux neutron facility for most of the neutron applications.

Novel Fabrication Routes of Metallic Micromembranes for In Situ Mechanical Testing

A methodology to miniaturize mechanical tests of metal alloys based on membrane deformation was developed in this investigation. The buildup of this new path for miniaturization tests requires small amounts of material for testing. This is of particular interest for irradiated structural nuclear materials. Micrometric metallic circular membranes were fabricated starting from thin alloy foils and using two different paths. Serial fabrication of microspecimens was performed by means of successive focused ion beam (FIB) steps. On the other hand, high-throughput parallel fabrication was achieved by differential sputtering (DS) based on reactive ion etching followed by a final fine FIB polishing to flatten the membranes and straighten the mechanical response. Micro-punch tests were performed using spherical tips and the in situ load–displacement curves were recorded while monitoring the test in a scanning electron microscope. The values reached after testing of the DS membranes were more reliable than those of FIB samples, showing a large stretching section and higher values of maximum force (64 mN) and displacement (22.2 μm). The micro-punch testing methodology developed in this work combines the advantage of facilitating the interpretation of the mechanical response, by producing a bi-axial stress distribution during membrane stretching, while being amenable to high-throughput microspecimen fabrication.

Helium-Hydrogen Synergistic Effects in Structural Materials Under Fusion Neutron Irradiation

In fusion reactors, 14 MeV high-energy neutron irradiation of structural materials will produce large amounts of helium and hydrogen simultaneously with displacement defects. These He and H atoms will interact with displacement defects, leading to He-H synergistic effects and aggravating the irradiation damage. Currently, there exist no available high-flux fusion neutron sources. Additionally, the neutron energy spectrum and the generation of He and H in fission reactors or spallation neutron sources greatly differ from those in fusion reactors. Multi-ion beam irradiation is a promising method to emulate the synergistic effects induced by fusion neutron irradiation. This review summarizes the experimental studies on the He-H synergistic effects, and analyzes the effects of He and H on cavity evolution and swelling under multi-ion beam irradiation. The roles of various experimental factors are also discussed. More systematically controlled experiments are suggested to develop a comprehensive understanding of He-H synergistic effects in structural materials.

Determination of the long-lived 10Be in construction materials of nuclear power plants using photoactivation method

The problem of handling nuclear power plant irradiated structural materials holds one of the central places in the nuclear power industry. High toxic 10Be with a half-life of T1/2 = 1.6 × 106 years is discovered in NPP structural materials after reactor operating. 10Be decays through only electrons' emission. Pure beta emitters are extremely difficult to determine in irradiated structural materials and radioactive waste. We proposed a photoactivation approach for determining the 10Be activity in NPP samples. The proposed method involves determining 9Be and 10B concentrations and subsequent recalculation of 10Be activity formed in 9Be(n, γ)10Be and 10B(n, p)10Be reactions. The amount of 9Be and 10B is determined by samples' photoactivation using an electron accelerator and 9Be(γ, 2n)7Be-, 10B(γ, p2n)7Be-reactions. These reactions' experimental yields were measured for 20, 40, and 55 MeV boundary energies of the bremsstrahlung beam. The proposed technique was tested on samples of ChNPP 2nd unit irradiated structural materials. The technique's calculated error is about 15–20%; the sensitivity is 1 Bq × g−1.

Ion irradiation induced changes in defects of iron thin films: Electron microscopy and positron annihilation spectroscopy

Single crystal Fe thin films (∼250 nm) were grown on MgO substrates and irradiated with 2.0 MeV Fe+ ions at 10 and 50 dpa, and the defect evolution was studied using high resolution Transmission Electron Microscopy (HR-TEM) and Doppler Broadening Positron Annihilation Spectroscopy (PAS). It was shown that irradiation induced or exacerbated a thin oxide layer at the outer interface and produced substantial Fe/Mg mixing at the film/substrate interface, particularly for the higher dose. Modeling of the PAS data allowed interpretation of the defect types at different distances from the Fe surface, and included several types of MgO substrate damage and annihilation condition changes, indicative of damage due to ballistic effects of the Fe atoms as well as chemical changes due to implantation and subsequent diffusion. This detailed PAS study compared with TEM and energy dispersive spectroscopy (EDS) provides significant insight into depth-dependent defect creation. These results will be useful for predicting defect creation in Fe-based materials under irradiation conditions, for extension to neutron irradiated structural materials.

Neutron-induced damage simulations: Beyond defect production cross-section, displacement per atom and iron-based metrics

Nuclear interactions can be the source of atomic displacement and post-short-term cascade annealing defects in irradiated structural materials. Such quantities are derived from, or can be correlated to, nuclear kinematic simulations of primary atomic energy distributions spectra and the quantification of the numbers of secondary defects produced per primary as a function of the available recoils, residual and emitted, energies. Recoils kinematics of neutral, residual, charged and multi-particle emissions are now more rigorously treated based on modern, complete and enhanced nuclear data parsed in state of the art processing tools. Defect production metrics are the starting point in this complex problem of correlating and simulating the behaviour of materials under irradiation, as direct measurements are extremely improbable. The multi-scale dimensions (nuclear-atomic-molecular-material) of the simulation process is tackled from the Fermi gradation to provide the atomic- and meso-scale dimensions with better metrics relying upon a deeper understanding and modelling capabilities of the nuclear level. Detailed, segregated primary knock-on-atom metrics are now available as the starting point of further simulation processes of isolated and clustered defects in material lattices. This allows more materials, incident energy ranges and particles, and irradiations conditions to be explored, with sufficient data to adequately cover both standard applications and novel ones, such as advanced-fission, accelerators, nuclear medicine, space and fusion. This paper reviews the theory, describes the latest methodologies and metrics, and provides recommendations for standard and novel approaches.

STUDYING THE CHANGES IN THE MECHANICAL PROPERTIES OF NEUTRON-IRRADIATED CHROMIUM-NICKEL STEELS USING THE “SHEAR PUNCH” METHOD

Obtaining mechanical properties of neutron irradiated structural materials is made difficult by the high radioactivity of typical specimen sizes. Developing new mechanical tests which obtain the same properties with far smaller specimen sizes would significantly enhance the speed and safety of performing such tests. In this study, we demonstrate the correspondence in mechanical properties between standard, uniaxial tensile testing and a new “Shear Punch” method, involving the shear-based removal of a small (1 mm) disc of material. Two types of steel, 12Cr18Ni9 and 08Cr16Ni11М3, were irradiated with fast neutrons in the BN-350 reactor up to 23 dpa and shear punch tested between 20…300 °C. Excellent correspondence was found between the 0.2% shear and uniaxial tensile yield stresses, validating this technique as a way to obtain the same mechanical properties with greatly reduced sample size.

Mesoscale modeling of irradiation damage evolution in bcc iron and vanadium: A comparative study

Voids and dislocation loops are two major types of damages in irradiated structural materials, which are mainly responsible for the degradation of material properties. Here we use the phase-field model and rate theory to simulate the microstructural evolution of voids and dislocation loops, respectively, in irradiated bcc iron and vanadium. The temperature-dependent material parameters of iron and vanadium are derived from ab initio calculations. The simulated results at different temperatures (513 K, 623 K and 722 K) and irradiation doses (1∼20 dpa) are analyzed to reveal the impact of irradiation conditions on the formation of irradiation-induced defect clusters. A comparison of the results shows larger void porosity and void/loop size in iron and higher void/loop density in vanadium. Then, a dispersed-barrier hardening model is used to correlate the mesoscale simulation results on microstructure with the yield stress change of the materials.

Irradiated FA Structural Materials Conditioning by Induction-Slag Remelting in a Cold Crucible

A method of managing one type of radwaste forming during BREST-300 reactor operation – the structural materials of FA – is described. In the module developed for reprocessing, the fissile materials losses together with all forms of waste must not exceed 0.1% of the initial amount, in connection with which the maximum admissible content of facade materials and structural materials sent into long-term storage must not exceed 0.001% by weight. Only the plutonium content may reach 0.015% by weight, which makes it necessary to perform additional purification and return fissile materials into the nuclear fuel cycle. A method of reprocessing the structural materials of irradiated fuel elements is proposed on the basis of induction-slag remelting in a cold crucible, which makes it possible to conduct purification to the desired index value. The operating regime of the melting unit and the qualitative composition of the employed flux that make it possible to reach the required level of purification by removal of fissile materials are described.

On the Problem of Modeling of Nuclear Transmutation Effects upon Investigation of the Phase Composition of Irradiated Austenitic Steels

In the problem of modeling of radiation-induced instability of MnS precipitates based on nuclear transmutation effects upon reactor irradiation, existing mathematical models have been considered in order to determine the design estimates of radiation-induced changes in the phase composition of irradiated structural materials of operating power and research reactors.

Nucleation of Pre-precipitates in Structural Materials under Cascade-Forming Irradiation

A model is proposed to calculate pre-precipitate nucleation rate in supersaturated solid solutions under cascade-forming irradiation. Estimates of the rate of nucleation of copper-enriched clusters in Fe–Cu solid solutions with a copper concentration of 0.05–0.18 at. % due to reactor irradiation are carried out. A comparison with the experimental data on irradiation of structural materials shows that the formation of precipitate nuclei takes place in the cascade region involved at the final stages of cascade relaxation, whose volume is about two times higher than the cascade volume at the dynamic stage.

Responses of Single-Cell and Differential Calorimeters: from Out-of-Pile Calibration to Irradiation Campaigns.

The nuclear radiation energy deposition rate (unit usually employed: $\mathrm {W.g}^{-1}$ ) is a key parameter for the thermal design of experiments on materials and nuclear fuel carried out in experimental channels of irradiation reactors, such as the French reactor in Saclay called OSIRIS or the Polish reactor named MARIA. In particular the quantification of nuclear heating allows the prediction of heat and thermal conditions induced in irradiated devices and/or structural materials. Various sensors are used to quantify this parameter, in particular radiometric calorimeters, also known as in-pile calorimeters. Two main kinds of in-pile calorimeter exist possessing two geometries and two measurement principles: the single-cell calorimeter and the differential calorimeter. The present work focuses on specific examples of these calorimeter types, from the step of their out-of-pile calibration (transient and steady experiments respectively) to the comparison between numerical and experimental results obtained from two irradiation campaigns (French and Polish reactors). The main aim of this paper is to propose a steady numerical approach to estimate the single-cell calorimeter response under irradiation conditions.

Modeling defect cluster evolution in irradiated structural materials: Focus on comparing to high-resolution experimental characterization studies

Abstract

Converting energy from fusion into useful forms

If fusion power reactors are to be feasible, it will still be necessary to convert the energy of the nuclear reaction into usable form. The heat produced will be removed from the reactor core by a primary coolant, which might be water, helium, molten lithium-lead, molten lithium-containing salt, or CO2. The heat could then be transferred to a conventional Rankine cycle or Brayton (gas turbine) cycle. Alternatively, it could be used for thermochemical processes such as producing hydrogen or other transport fuels. Fusion presents new problems because of the high energy neutrons released. These affect the selection of materials and the operating temperature, ultimately determining the choice of coolant and working cycle. The limited temperature ranges allowed by present day irradiated structural materials, combined with the large internal power demand of the plant, will limit the overall thermal efficiency. The operating conditions of the fusion power source, the materials, coolant, and energy conversion system will all need to be closely integrated.

Simulation Experiment on Study in the Radiation Resistance of Advanced Ferrite-Martensite Steel Hardened by Disperse Inclusions

It is shown that simulation experiments on the heavy-ion irradiation of structural materials used in nuclear power facilities can be used for tomographic atomic-stand studies. Experiments have been performed on irradiation of samples of advanced steel used in nuclear and fusion reactors EK-181 and ODS Eurofer by iron ions to damaging dose 10 dpa. The steel is characterized by the presence of a large number of nanosize (2–3 nm) clusters, dispersion-hardening these materials. Irradiation was performed on the TIPr-1 heavyion linear accelerator. Atomic-probe studies of irradiated samples reveal a change in the composition of nanosize clusters under irradiation.

Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 °C and 500 °C. During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 °C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 °C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.