Irradiated Steel Research Articles

Carbon-14 Release and Speciation During Corrosion of Irradiated Steel Under Radioactive Waste Disposal Conditions

Carbon-14 Release and Speciation During Corrosion of Irradiated Steel Under Radioactive Waste Disposal Conditions

COMPARISON OF WASTE DUE TO IRRADIATED STEELS IN THE ESFR AND DEMO

For either nuclear fusion or generation IV fission reactors to be viable as a commercial energy source the decommissioning and waste disposal solutions must be considered during the design. A multi-step simulation process combining Monte Carlo Neutron Transport simulations with inventory simulations have been performed to estimate the activation of steels in key reactor components of the European Sodium-cooled fast Reactor (ESFR). Waste classifications based on UK waste disposal regulations have been applied to the key components to estimate the expected masses of low level and intermediate level waste. The use of reduced activation steels, EUROFER and F82H, in reactor components external to the core results in a factor of 10 reduction in the percentage of waste classified as Intermediate Level Waste (ILW). Waste estimates are compared to existing waste estimates for the European Demonstration fusion reactor (DEMO). The ESFR has a lower percentage of ILW per total reactor mass due to irradiated steels compared to DEMO. However, there is no Higher Activity Waste (HAW) associated with DEMO, compared with arisings from the ESFR spent fission fuel.

Effect of helium concentration on irradiation damage of Fe-ion irradiated SIMP steel at 300 °C and 450 °C* *Project supported by the National Natural Science Foundation of China (Grant Nos. U1832133 and 12075194), Sichuan Provial Science and Technology Program, China (Grant No. 2020ZYD055), and the Doctor Research Foundation of Southwest University of Science and Technology (Grant No. 18zx714101).

SIMP steel is newly developed fully martensitic steel for lead-cooled fast reactors and accelerator-driven systems. It is important to evaluate its radiation resistance via high flux neutron irradiation, where dense He atoms can be formed via (n, α) transmutation reaction. Co-irradiation with Fe and He ions, instead of neutron, was performed. Specimens were irradiated with 6.4-MeV Fe ions to the damage dose of 5 dpa at a depth of 600 nm. Three different helium injection ratios of 60-appm He/dpa (dpa: displacements per atom), 200-appm He/dpa and 600-appm He/dpa at a depth of 600 nm, were performed. Two different irradiation temperatures of 300 °C and 450 °C were carried out. The effect of helium concentration on the microstructure of Fe-irradiated SIMP steel was investigated. Microstructural damage was observed using transmission electron microscopy. The formed dislocation loops and bubbles depended on the helium injection ratio and irradiation temperature. Lots of dislocation loops and helium bubbles were homogeneously distributed at 300 °C, but not at 450 °C. The causes of observed effects are discussed.

Characterization of mechanical properties and deformation behavior of highly irradiated 316L stainless steel from target modules at the Spallation Neutron Source using digital image correlation analysis

Tensile specimens with irradiation doses ranging from approximately 0.7 to 7.6 displacements per atom (dpa) were fabricated from 316L samples removed from Spallation Neutron Source (SNS) target vessels after service. Tensile tests were performed in conjunction with digital image correlation (DIC) characterization, which permitted detailed analysis of the strain distribution and deformation behavior of the specimen gauge sections during testing. The results show, while considerable hardening (500–700 MPa increase in yield strength) occurred in the 316L SNS target material with increasing dose, appreciable ductility (19–23% total elongation) remained in the material after irradiation to 7.6 dpa in the mixed neutron/proton environment of the SNS target region. The DIC data showed the strain distribution and deformation behavior of all irradiated specimens were nonuniform during testing. Strain localization in the form of deformation bands or “hotspots” was observed during testing, and in some specimens, the bands propagated as deformation waves along the gauge section. The “traveling wave” deformation behavior affected the amount of elongation each specimen achieved, and the scatter in the elongation data was attributed to the large variation in deformation behavior observed among different specimens. Bands of localized strain, or deformation bands, oriented at specific angles were observed in two-dimensional plots of the strain distribution in the gauge sections of some specimens. This paper describes the experimental procedures, results, and deformation behavior of tensile specimens from SNS targets after service and illustrates the value of DIC characterization during tensile testing of irradiated materials.

Generation of particles and fragments by quasicontinuous wave fiber laser irradiation of stainless steel, alumina, and concrete materials

An experimental characterization is presented on fine particles, droplets, and fragments produced at the interaction region between a 2.7 kW quasicontinuous wave (repetitive pulsed operation with a 10 ms pulse duration) fiber laser and stainless steel, alumina, and heavy concrete samples. In the samples, the recoil pressure induced by vaporization pushes particles and fragments into the ambient atmosphere. In order to preserve a safe working environment, in particular for nuclear decommissioning, special care should be taken to confine and retrieve such particles during laser processing. In the experiments, particle production from the vapor and the molten phase layer in the targeted material was observed with a high-speed camera, with fine particles collected and analyzed using an electron microscope. The observed results were qualitatively interpreted with the help of a simplified one-dimensional hydrodynamic code coupled with a stress computation code. Characterization and classification of the results are expected to provide a useful database that will contribute to the decommissioning of nuclear facilities and other industrial applications.

Method of Convex Marking of the Surfaces of Steel Products Using a Pulsed 50-W Infrared Fiber Laser

Convex linear structures of various configurations with a height up to 200 μm and a depth up to 100 μm and point structures with a height up to 500 are formed on the surface of steel samples using a 50-W pulsed nanosecond fiber laser with a wavelength of 1.064 μm without filler wires or powders. The structures are formed from the irradiated steel material by controlling the melt dynamics. The possibility of forming braille on a steel surface using this technology is shown. The results are promising for a number of industries, where electron-beam machines and high-power laser sources are prohibitively expensive and a high rate of relief formation is required.

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.

A micromechanical analysis of intergranular stress corrosion cracking of an irradiated austenitic stainless steel

Irradiation Assisted Stress Corrosion Cracking (IASCC) is a material degradation phenomenon affecting austenitic stainless steels used in nuclear Pressurized Water Reactors (PWR), leading to the initiation and propagation of intergranular cracks. Such phenomenon belongs to the broader class of InterGranular Stress Corrosion Cracking (IGSCC). A micromechanical analysis of IGSCC of an irradiated austenitic stainless steel is performed in this study to assess local cracking conditions. A 304L proton irradiated sample tested in PWR environment and showing intergranular cracking is investigated. Serial sectioning, Electron BackScatter Diffraction (EBSD) and a two-step misalignment procedure are performed to reconstruct the 3D microstructure over an extended volume, to assess statistically cracking criteria. A methodology is also developed to compute Grain Boundary (GB) normal orientations based on the EBSD measurements. The statistical analysis shows that cracking occurs preferentially for GB normals aligned with the mechanical loading axis, but also for low values of the Luster-Morris slip transmission parameter. Micromechanical simulations based on the reconstructed 3D microstructure, FFT-based solver and crystal plasticity constitutive equations modified to account for slip transmission at grain boundaries are finally performed. These simulations rationalize the correlation obtained experimentally into a single stress-based criterion. The actual strengths and weaknesses of such micromechanical approach are discussed.

Concomitant oxidation-diffusion-creep processes for stress generation and suppression of oxide-alloy interfacial instabilities

Metallic materials under extreme thermo-mechanical-environmental service conditions oftentimes rely on the formation of dense and protective oxide layers, e.g., typically alumina or chromia. Their failures, however, usually emerge from the morphological instability of such layers at the oxide-alloy interface. This work suggests that a tensile, biaxial stress arises in the substrate alloy due to the concomitant oxidation-diffusion-creep processes, which, in contrast to the growth stress in the oxide layer, provides a stabilizing factor that suppresses the rumpling or compositional variation in chromia or alumina interlayers. For rumpling analysis, the substrate tensile stress compensates the growth stress in the oxide layer. The compositional variation is modeled by the stress-domain analysis, which suggests that the substrate tensile stress prefers the growth of high-growth-stress oxide phase at the expense of low-growth-stress one. Our findings have been successfully employed to rationalize recent experiments in superalloys under cyclic oxidation conditions and in carburized or irradiated stainless steels under isothermal oxidation conditions.

Atom probe characterisation of segregation driven Cu and Mn–Ni–Si co-precipitation in neutron irradiated T91 tempered-martensitic steel

The T91 grade and similar 9Cr tempered-martensitic steels (also known as ferritic-martensitic) are leading candidate structural alloys for fast fission nuclear and fusion power reactors. At low temperatures (300–400 °C) neutron irradiation hardens and embrittles these steels, therefore it is important to investigate the origin of this mode of life limiting property degradation. T91 steel specimens were separately neutron irradiated to 2.14 dpa at 327 °C and 8.82 dpa at 377 °C in the Idaho National Laboratory Advanced Test Reactor. Atom probe tomography was used to investigate the segregation driven formation of Mn–Ni–Si-rich (MNSPs) and Cu-rich (CRP) co-precipitates. The precipitates increase in size and, slightly, in volume fraction at the higher irradiation temperature and dose, while their corresponding compositions were very similar, falling near the Si(Mn,Ni) phase field in the Mn–Ni–Si projection of the Fe-based quaternary phase diagram. While the structure of the precipitates has not been characterised, this composition range is distinctly different than that of the typically cited G-phase. The precipitates are composed of CRP with MNSP appendages. Such features are often observed in neutron irradiated reactor pressure vessel (RPV) steels. However, the Si, Ni, Mn, P and Cu solutes concentrations are lower in the T91 than in typical RPV steels. Thus, in T91 precipitation primarily takes place in solute segregated regions of line and loop dislocations. These results are consistent with the model for radiation induced segregation driven precipitation of MNSPs proposed by Ke et al. Cr-rich alpha prime (α’) phase formation was not observed.

Wear Resistance Structural Aspects of Materials after Laser Processing

The conducted experiments show that the structures generated by laser processing are able to dissipate friction energy effectively with transformations at different structure levels. Thus, it becomes possible to manage surface strength and wear resistance of materials using the concept of “structural adaptability” of friction pairs that expands the range of normal friction and wear processes. The irradiated steel wear resistance is defined by its original hardness as well as its ability to deformation strengthening during friction. The temperature and strength load of irradiated steel surface layers through friction destabilize austenite towards y-xx transformation. This results from alloy carbide formation which causes depleted austenite by carbon, and contributing to its transformation into martensitic deformation. It’s necessary to regulate the amount and degree of remaining austenite stability in laser-hardened steel and alloys, applied to specific load conditions, that provides that required operation features.

A Study on the Removal Characteristics of a Radioactively Contaminated Oxide Film from the irradiated Stainless Steel Surface using Short Pulsed Laser Ablation

A Study on the Removal Characteristics of a Radioactively Contaminated Oxide Film from the irradiated Stainless Steel Surface using Short Pulsed Laser Ablation

Hardening behavior and deformation microstructure beneath indentation in heavy ion irradiated 12Cr-ODS steel at elevated temperature

Hardness behavior and deformation of the microstructure beneath the indentation in 12Cr-ODS steel after exposure to heavy ion irradiation at 573 K were investigated with combined applications of nano-indentation tests, EBSD analysis, and TEM observation. The inhomogeneous nano-indentation hardness behavior was firstly investigated in both non-irradiated and irradiated specimens and compared to those at room temperature (RT) irradiation. Subsequently, TEM observations were performed to understand the deformation behaviors beneath the indentation tested at typical low-indexed orientations. Obvious irradiation hardening was confirmed in both the normal direction (ND) and rolling direction (RD) specimens, and the hardening effect intensified with an increase in irradiation dose. Meanwhile, various anisotropic hardness behaviors were observed such as a lower hardness value and a weaker hardening effect in the RD specimen compared to the ND specimen, and a lower hardness value but a greater hardening effect in [001] orientation relative to [111] orientation. Furthermore, results from 573 K irradiation showed similar anisotropic hardness behaviors with those obtained from RT irradiation, despite the fact that irradiation at 573 K exhibited a weakened irradiation hardening effect in this steel. TEM observation of the deformation microstructure beneath the indentation reveals that (i) the depth of the deformation affected zone varied with orientation, ranging from ~6-9 times that of the indent depth, and (ii) distinct dislocation-grain boundary (GB) interactions beneath the indentations was confirmed to exhibit dependence on the geometric relation between GB and indenter impression direction. Correlation between the anisotropic hardness behavior and microstructural features is discussed based on Schmid factor analysis, comparison of geometric favorability between indenter and possible slip planes, and TEM observation of deformation microstructure. The results conclude that the anisotropic hardness behaviors in the present 12Cr-ODS steel are ascribed to the combined effect of crystal orientation and the non-uniform grain morphology with the existence of the elongated grains having low aspect ratio in the ND plane and nano-layered grains in the RD plane.

Radiation Damage of Reactor Pressure Vessel Steels Studied by Positron Annihilation Spectroscopy—A Review

Safe and long term operation of nuclear reactors is one of the most discussed challenges in nuclear power engineering. The radiation degradation of nuclear design materials limits the operational lifetime of all nuclear installations or at least decreases its safety margin. This paper is a review of experimental PALS/PLEPS studies of different nuclear reactor pressure vessel (RPV) steels investigated over last twenty years in our laboratories. Positron annihilation lifetime spectroscopy (PALS) via its characteristics (lifetimes of positrons and their intensities) provides useful information about type and density of radiation induced defects. The new results obtained on neutron-irradiated and hydrogen ions implanted German steels were compared to those from the previous studies with the aim to evaluate different processes (neutron flux/fluence, thermal treatment or content of selected alloying elements) to the microstructural changes of neutron irradiated RPV steel specimens. The possibility of substitution of neutron treatment (connected to new defects creation) via hydrogen ions implantation was analyzed as well. The same materials exposed to comparable displacement damage (dpa) introduced by neutrons and accelerated hydrogen ions shown that in the results interpretation the effect of hydrogen as a vacancy-stabilizing gas must be considered, too. This approach could contribute to future studies of nuclear fission/fusion design steels treated by high levels of neutron irradiation.

Irradiation damages of structural materials under different irradiation environments

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study's Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

Развитие комплексного электронно-ионно-плазменного метода модификации поверхности материалов и изделий

The aim of the research is to elucidate the patterns of modification of the elemental and phase composition, defective substructure, and mechanical and tribological properties of a carbon steel during a combined treatment in a single vacuum cycle, including deposition of niobium and silicon films on the steel surface and irradiation of the film/substrate system with an intense pulsed electron beam. The formation in a single vacuum cycle of Si–Nb films 1 μm thick each on the 40Cr steel specimens was carried out on the COMPLEX setup by the magnetron (Si) and arc evaporation (Nb) methods with plasma assistance. High-speed melting of the (Si–Nb) film/(40Cr steel) substrate system was carried out by an intense pulsed electron beam. The mode of irradiation of the (Si + Nb) film/(40Cr steel) substrate system by an intense pulsed electron beam (50 J/cm2, 200 μs, and 3 pulses) is revealed that allows the formation of the surface layer characterized by the high hardness equal to 9600 MPa (exceeding the hardness of the 40Cr steel in the initial (ferrite-perlite) state by a factor of 3.3) and the wear resistance that exceeds by more than 15 times the wear resistance of the 40Cr steel in the initial state. It has been shown that high strength and tribological properties of the steel are due to the formation of particles of the strengthening phase (Nb5Si3, Nb3.61Si0.39, Nb08Fe02, SiC, and NbO2) and a relatively large (up to 30 μm) thickness of the strengthened layer.

FRACTOGRAPHY OF IRRADIATED STEEL 0.12C18Cr10NiTi IN A BN-350 REACTOR AFTER MECHANICAL TESTS AT TEMPERATURES OF 350 AND 450 °C

The results are presented of a fractographic study of the fracture surface of samples of irradiated steel 0.12C18Cr10NiTi after tensile tests at temperatures of 24, 350, 450 °C. Samples for research were cut from the face of duct of spent fuel assemblies of fast neutron reactor BN-350. On the basis of a multilevel approach of physical mesomechanics, an attempt is made to explain the decrease in the ductility of steel with an increase in the temperature of mechanical tests. Based on the analysis results, it was concluded that a decrease in the ductility of steel with an increase in the test temperature may be due to quasi-uniform stress distribution in the area of localization of deformations, which was facilitated by the local increase in the porosity of the material caused by the accommodative processes of the rotary type.

Technical Basis Part 2 for Code Case N-889: Reference Stress Corrosion Crack Growth Rate Curves for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments that were adopted as ASME B&PV Section XI Code Case N-889. As described in a technical basis Part 1 paper, over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC) and pressurized water reactor (PWR) primary water environments. Part 1 also describes the custom fitting process, quality of fit, and comparisons with related literature and data not used for fitting. This technical basis Part 2 paper describes shifting the mean models to the 75th percentile of the calibration data, simplifying to produce the N-889 curves, and comparing with previous reference curves and over 500 data points not used for developing the N-889 curves, including weld, cast, and heat-affected-zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations using both yield stress and CGR equations.

Technical Basis Part 1 for Code Case N-889: Data Selection and Modeling of Stress Corrosion Crack Growth Rates for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) models and reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments. The curves have been approved as ASME B&PV Section XI Code Case N-889, and this paper is the technical basis Part 1 for Case N-889, summarizing collection, review, and selection of calibration data, modeling, and analysis of fit. Over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC), and pressurized water reactor (PWR) primary water environments. This Part 1 paper also discusses related literature and compares with other data not used for mean model calibration. A technical basis Part 2 paper describes shifting the mean models to the 75th percentile, simplifying to produce Case N-889 reference curves, and comparing with previous reference curves and over 500 data points not used for reference curve development, including weld, cast, and heat affected zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations.

Crack growth prediction for irradiated stainless steels under the combined fatigue-creep loading

A methodology is presented in this study for estimating the crack growth of austenitic stainless steel Types 304, 304L, 316, and 316L, which are commonly used as structural materials in nuclear pressure vessels. These structural components are typically subjected to neutron irradiation and combined loadings, including repeated mechanical stresses (i.e., fatigue) resulting from startups and shutdowns, as well as creep due to on-load periods at elevated temperature. In this study, the fatigue crack length was estimated using a strip-yield based fatigue crack growth model. The model is extended to include the effects of creep deformation in a presence of hold- times, and expanded to include the effects of irradiation. Reasonable crack growth estimations are obtained for selected materials under various combined loading conditions when compared to the experimental data available in the literature.