Irradiated Steel Research Articles

Study of defect complexes and their evolution with temperature in hydrogen and helium irradiated RAFM steel using positron annihilation spectroscopy

The formation of hydrogen and helium associated defect complexes and their influence on the nucleation and growth of helium bubbles in Indian Reduced Activation Ferritic/Martensitic steel have been investigated using positron annihilation spectroscopy. Isochronal annealing study has been performed on the samples individually irradiated with H and He ions, and sequentially irradiated with both H and He ions. All the as-irradiated samples show higher S-parameter as compared to the unirradiated sample due to the presence of irradiation-induced vacancy-type defects. Among the single ion irradiated samples, hydrogen irradiated sample shows the presence of H-vacancy complexes having high hydrogen content in the as-irradiated state, an increase in S-parameter due to the release of hydrogen from these complexes at 373 K, and complete defect recovery after annealing at 673 K. The helium irradiated sample shows three distinct stages of annealing; a defect annealing stage with decrease in S-parameter from as-irradiated to 573 K, bubble nucleation stage with a stable S-parameter from 573 to 673 K and the bubble growth region characterized by an increase in S-parameter is observed from 673 to 973 K. Among the sequential irradiated samples, the sample irradiated with helium first and then with hydrogen shows a reduced S-parameter which is due to the saturation of helium-induced vacancies by hydrogen or the formation of H-He-vacancy complexes. Both the sequential irradiated samples show a reduced S-parameter from 573 to 673 K due to the formation of H-He-vacancy complexes under post-irradiation annealing. The nucleation and growth stage of helium bubbles observed under isochronal annealing is independent of the sequence of irradiation. Both the sequential irradiated samples show the bubble nucleation and growth stage similar to helium irradiated sample and further, the presence of hydrogen has no noticeable role in this stage.

Effect of Nitrogen Implantation on the Structure and Properties of Austenitic Corrosion-Resistant Steels

Work is devoted to studying the effect of implantation of nitrogen ions into the surfaceof austenitic stainless steels to improve their functional properties. Four grades ofaustenitic corrosion-resistant steels 02H16N10M2, 08H15AG10D2, 06H15AG9NM2 and09H15AG9ND2 were taken after cold plastic deformation and annealing from 680 ∘Cin water and subsequent implantation with N+ ions with different radiation dose: 0,01 и0,1%. It was found that irradiation of austenitic stainless steels with nitrogen ions can beconsidered an effective way to increase the hardness and yield strength of steels in theoperation in a corrosive environment.

Influence of Neutron and Electron Irradiation on Structural-Phase Transformations in Fe–12Cr–2W–V–Ta–B Steel Processed under Various Heat Treatment Conditions

The effect of electron and neutron irradiation and post-radiation annealing on the properties of 12% chromium ferritic-martensitic low-activated EK-181 steel treated under various heat conditions has been studied using transmission electron microscopy and resistometry. Both electron and neutron irradiation of EK-181 steel in the 300–340 K temperature range result in radiation-induced solid solution separation during all subsequent heat treatments. Isochronous annealing in the 300–500 K temperature range causes additional solid solution separation. Nanoclusters of point defects with a size from 1.5 to 5 nm and a concentration of 5 × 1016 cm–3 have been observed after neutron irradiation at a fluence of (1–5) × 1019 cm–2. Thermal homogenization of the solid solution occurs above 650 K.

Resistivity Recovery Stages of Proton Irradiated EUROFER97 Steel

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Use of Irradiated Fracture Toughness Values in Nuclear Vessel and Component Design Specifications for Fitness-For-Service Analysis Using the Unified Curve Method

The American Society of Mechanical Engineers (ASME) Section III Rules for Construction of Nuclear Facility Components subsection NB-2331 Material for Vessels requires that the effects of irradiation shall be considered on material toughness properties in the core belt line region of the reactor vessel. The code also states that “the design specifications shall include additional requirements, as necessary, to ensure adequate fracture toughness for the service lifetime of the vessel.” In a design report of nuclear pressure vessel, the design and service loads do not include loads that are affected by fracture toughness of the material. However, in the cases of fitness-for-service assessment for component flaws (prevalent with age of component), irradiated material properties become highly relevant. An example of a fitness-for-service is that of a beyond design basis reactor vessel head drop accident in a pressurized water reactor with a nozzle junction flaw. As a case study, the critical size of a postulated external surface semielliptical circumferential crack in the combustion engineering three-loop pressurized water reactor nozzle–vessel junction is calculated using ansys Workbench (Academic version) with the applied impact load from the vessel head drop accident. Failure assessment diagrams for numerous crack depths and lengths were developed considering the fracture toughness properties of the irradiated reactor vessel steel. The mode I stress intensity results used in the failure assessment diagram were compared with the available finite element and the American Petroleum Institute (API) standard API 579 analytical solutions for validation, showing good agreement. From this case study, it is demonstrated that the effects of irradiation on fracture toughness become prominent at the same postulated crack size in the nozzle–vessel junction dispositioned as “safe” becomes “unsafe” in the fracture assessment diagram. Using the unified curve method, the irradiated fracture toughness data in design specification can be supplied so that it may be used in fitness-for-service analysis to account for component aging.

Prediction of irradiation hardening in austenitic stainless steels: Analytical and crystal plasticity studies

The flow stress in irradiated Austenitic Stainless Steels (ASSs) is analyzed in terms of the microstructure features and the fundamental mechanisms of plastic deformation. A set of constitutive equations is proposed to assess the contributions of dislocation network, grain size effect, dislocation loops and solute clusters, accompanied with an appropriate superposition rule. Using simple equations for the evolution of densities of dislocations and radiation defects, it is possible to integrate numerically the tensile stress-strain curves over a wide range of irradiation doses. It is shown that irradiation hardening can be obtained analytically without adjustable parameters. Simulation results are challenged in comparison with experimental results and found in excellent agreement with all the reported experimental trends in ASSs irradiated at conditions close to those prevailing in pressurized water reactors.

Quantitative linkage between the stress at dislocation channel – Grain boundary interaction sites and irradiation assisted stress corrosion crack initiation

Localized deformation has emerged as a key factor in the crack initiation process for irradiated steels, as cracks are observed to nucleate preferentially at these sites. Using high resolution electron backscatter diffraction (HREBSD), the local stress tensor surrounding the dislocation channel-grain boundary interaction sites was quantified and coupled with fully determined grain boundary plane orientation information to determine, for the first time, the relationship between grain boundary normal stress and intergranular crack initiation in irradiated austenitic stainless steel. A Fe13Cr15Ni alloy was strained in simulated boiling water reactor, normal water chemistry after quantifying the residual stress tensor at discontinuous dislocation channel – grain boundary interaction sites where grain boundaries were determined to be well oriented with respect to the loading axis. Local stresses at the grain boundary were observed to reach magnitudes greater than 1.5 GPa at a distance of 200 nm from the intersection between the dislocation channel and the grain boundary. A pseudo-threshold stress of 0.9 GPa was measured, below which no cracking was observed. As the stress acting normal to the grain boundary increased above this value, the susceptibility to cracking increased with the cracking fraction reaching 100% at the high end of the stress range. This study shows for the first time that not only does intersection between discontinuous dislocation channels and grain boundaries result in peak local stresses, but the magnitude of the local tensile stress drives the crack initiation process.

Enhanced irradiation and corrosion resistance of 316LN stainless steel with high densities of dislocations and twins

The irradiation and corrosion damages of nuclear materials have been a major challenge for the nuclear energy industry. Here, rotationally accelerated shot peening was used to improve the radiation and corrosion resistance of 316LN austenitic stainless steel by introducing high densities of dislocations and nanoscale twin boundaries. The dislocations and twins not only impeded the formation of helium bubbles and shear bands during irradiation, but also facilitated the formation of a superior passivation film, which significantly improved its corrosion performance. These observations provide an effective approach for designing irradiation- and corrosion-resistant nuclear materials.

High-pressure torsion assisted segregation and precipitation in a Fe-18Cr-8Ni austenitic stainless steel

During irradiation or high-temperature aging, stainless steel can develop precipitates, significantly affecting mechanical properties. In this study atom probe tomography (APT) was used to study grain boundary segregation and secondary phases in a purely austenitic SS304 (a Fe-18Cr-8Ni steel) processed by high-pressure torsion (HPT) at 300 °C. Ni, Mn, Si enriched phase was observed at grain boundaries, and Cu nanoprecipitates were observed along and near phase/grain boundaries. Precipitation is facilitated by deformation assisted segregation along grain boundaries with a mechanism similar to vacancy diffusion in irradiated steels.

Austenite-based Stainless Steel Irradiation Behavior of the Precipitate and Void Swelling

Austenite-based Stainless Steel Irradiation Behavior of the Precipitate and Void Swelling

Radiation-hardening and nano-cluster formation in neutron-irradiated 9Cr[sbnd]2W low activation steels with different Si contents

Two kinds of 9Cr2W steels with and without the addition of 0.1 wt% Si were irradiated with neutrons up to a fluence of 4.8 × 1023 n/m2 at 563 K. Precipitates, dislocation loops, radiation-enhanced nano-clusters and radiation-hardening were studied using advanced electron microscopy and micro-hardness test. Si reduced the number density of dislocation loops, and therefore the extent of radiation-hardening decreased. It also favored the formation of silicide-like nano-clusters. In this study, the role of Si in neutron irradiated 9Cr2W steels is discussed, and then extended to nano-clusters and irradiation-hardening.

Grain orientation dependence of nanoindentation and deformation-induced martensitic phase transformation in neutron irradiated AISI 304L stainless steel

We studied the mechanical properties of a neutron irradiated AISI 304L stainless steel using nanoindentation with respect to crystallographic grain orientation. After neutron irradiation, {110} planes exhibit slightly larger increase in hardness and modulus compared to {100} and {111} planes. Nanoindentation-induced twinning occurs in the unirradiated specimen, while austenite-to-martensite phase transformation arises in its irradiated counterpart. Deformation-induced martensitic transformation is related to the Gibbs free energy associated with the irradiation-induced voids, compensating the surface energy necessary to create martensite laths.

Modeling the Influence of Strain Fields Around Precipitates on Defect Equilibria and Kinetics Under Irradiation in ODS Steels: A Multi Scale Approach

Resistance to radiation damage is a crucial property for structural materials in fission or fusion reactors. The particle-matrix interface in oxide dispersion strengthened (ODS) steels is considered a sink for point defects created during irradiation. In this work we address thequestion, if elastic strain fields around oxide particles cause a long-range interaction between the precipitates and point defects. We use kinetic Monte Carlo simulations to simulate the diffusion of point defects under the influence of elastic strain fields caused by Y2O3 and Y2Ti2O7 precipitates. We show that there is essentially no interaction between vacancies and the strain fields, while the sink strength of precipitates for interstitials increases with misfit strain between precipitate and matrix. Consequently, the interstitial concentration decreases with misfit strain, while due to reduced interstitial-vacancy recombination the vacancy concentration increases with misfit strain. The total change of point defect concentration with misfit strain is, however, rather limited.

Reverse evolution in nanoscale Cu-rich precipitates of an aged Fe–Cu alloy under electropulsing

ABSTRACTNanoscale Cu-rich precipitates (CRPs) are one of the most important microstructural nano-features responsible for embrittlement and hardening of reactor pressure vessels (RPVs), which threaten the safe operation of nuclear power plants (NPPs) and hinder the lifetime extension of nuclear reactors. A thermally aged Fe-1.1 wt.%Cu alloy, which is used to simulate embrittlement of the irradiated RPV steels, was treated by electropulsing with various parameters. The effect of electropulsing on nanoscale CRPs was investigated by using transmission electron microscopy (TEM). Compared to the traditional heat treatment, the electropulsing treatment (EPT) can accelerate the dissolution of CRPs in an aged Fe-Cu alloy on account of the higher atomic drift flux and the additional Gibbs free energy induced by electropulsing. More importantly, EPT is likely to be a new way of eliminating irradiation-induced Cu-rich precipitates.

Improving the functional properties of austenitic steels when modifying the surface with high-energy nitrogen ions

Abstract The work is devoted to the study of the influence of the implantation of nitrogen ions into the surface of austenitic corrosion-resistant steels in order to increase their functional properties, which is a relevant and significant task for practice. It has been established that irradiation of austenitic corrosion-resistant steels with nitrogen ions can be considered an effective way to increase hardness and the conditional yield strength in the operation of steels functioning in a corrosive environment.

14C Exposure from Disposal of Radioactive Waste Compared to14C Exposure from Cosmogenic Origin

ABSTRACTThe potential14C (carbon-14, radiocarbon) flux from disposal of14C containing waste into air is compared with the natural14C emanation rate from soil in order to put the14C hazard potential from disposal of this waste in perspective with the14C exposure from cosmogenic origin. Chemical corrosion of neutron irradiated metals, steel and Zircaloy, is bounded by diffusion of water through a thermodynamically stable metal-oxide layer and dissolution of this metal-oxide in a nuclear plant. Many countries process radioactive waste for disposal using cementitious materials, an acknowledged end-point management technique for this waste. The metal-oxides are also stable when these waste forms are embedded in cementitious materials. The14C release rate from this Zircaloy at these alkaline and reducing conditions is comparable to the natural14C emanation rate from soil into air. Neutron irradiated graphite and spent ion exchange resins are chemically inert and therefore other release mechanisms need to be assumed. Radiolytic corrosion is used to determine the14C release rate from this graphite. Moreover, ion exchange—with ingressing anionic species that have a higher affinity than contained anionic14C—is proposed as a release mechanism for these resins.

An approach in the analysis of microstructure of proton irradiated T91 through XRDLPA using synchrotron and laboratory source

The changes in the microstructure of 3.5 MeV proton irradiated T91 Ferritic-Martensitic steel samples with dose have been evaluated using detailed X-ray diffraction line profile analysis of the data collected using both laboratory and the synchrotron source. Different line profile analysis techniques like Williamson-Hall, modified Rietveld method and convolutional multiple whole profile fitting have been applied to evaluate the microstructural parameters such as domain size, microstrain, dislocation density and the character of the dislocation from both the data. The coherent domain size decreases and the corresponding microstrain increases at the first dose of irradiation. With further irradiation, these parameters show a slight recovery and eventually tend towards saturation with increasing dose. The dislocation density follows similar trend as the microstrain showing saturation at higher doses of irradiation in both the XRD data. However, a systematic change in the character of the dislocation from screw-type to edge-type could be ascertained only from the data obtained using synchrotron source. The Vickers microhardness of the samples is found to increase continuously with increasing irradiation dose supporting the observed change in the dislocation character from screw-type (glissile) to edge-type (sessile). Both EBSD and TEM analysis of the unirradiated and irradiated samples have supported the results obtained from the XRD analysis.

Identification of chemical form of stable carbon released from type 304L and 316L stainless-steel powders in alkaline and acidic solutions under low-oxygen conditions

ABSTRACTThe chemical form of14C released from irradiated stainless steel is a key parameter in the safety assessment of the subsurface disposal system in Japan. In this study, to identify the chemical form of the released carbon, unirradiated stainless-steel powders, which were found to be water-atomized powders with a silicon oxide film, were immersed in NaOH and HCl solutions under low-oxygen conditions for approximately 25 days. The results showed that the main chemical forms of the carbon were colloidal carbon in the NaOH solution and colloidal carbon and formic and acetic acids in the HCl solution. Almost no hydrocarbons were detected in both solution systems. Concerning the source of the colloidal carbon and carboxylic acids, the hypothesis that carbon in the oxide layer is released is considered to be reasonable. The very small amounts of hydrocarbons generated prevented us from discussing the source of the hydrocarbons. To validate the hypothesis and obtain further information on the hydrocarbons, additional experiments are necessary. In particular, for long-term safety assessment, it is important to determine whether the colloidal carbon, carboxylic acids and hydrocarbons are continuously released during the corrosion process. Therefore, information on the temporal evolution of the carbon should be obtained.

14C release from irradiated stainless steel

ABSTRACTRadiocarbon (14C or carbon-14, half-life 5730 yr) is a key radionuclide in the assessment of the safety of a geological disposal facility (GDF) for radioactive waste. In particular, the radiological impact of gaseous carbon-14 bearing species has been recognized as a potential issue. Irradiated steels are one of the main sources of carbon-14 in the United Kingdom’s radioactive waste inventory. However, there is considerable uncertainty about the chemical form(s) in which the carbon-14 will be released. The objective of the work was to measure the rate and speciation of carbon-14 release from irradiated 316L(N) stainless steel on leaching under high-pH anoxic conditions, representative of a cement-based near field for low-heat generating wastes. Periodic measurements of carbon-14 releases to both the gas phase and to solution were made in duplicate experiments over a period of up to 417 days. An initial fast release of carbon-14 from the surface of the steel is observed during the first week of leaching, followed by a drop in the rate of release at longer times. Carbon-14 is released primarily to the solution phase with differing fractions released to the gas phase in the two experiments: about 1% of the total release in one and 6% in the other. The predominant dissolved carbon-14 releases are in inorganic form (as 14C-carbonate) but also include organic species. The predominant gas-phase species are hydrocarbons with a smaller fraction of 14CO (which may include some volatile oxygen-containing carbon-species). The experiments are continuing, with final sampling and termination planned after leaching for a total of two years.

Mechanisms of radiation-induced segregation around He bubbles in a Fe-Cr-Ni crystal

Fe-15Cr-15Ni (wt.%) austenitic stainless steel was irradiated with 200 keV 4He+ ions to a fluence of 5 × 1016/cm2 at 500 °C. Radiation-induced segregation (RIS) at He bubbles in the irradiated steel was measured using Energy Dispersive X-ray Spectroscopy (EDS) coupled with Scanning Transmission Electron Microscopy (STEM). The RIS measurements exhibited an enrichment of Ni and a depletion of Cr and Fe at the bubbles. To determine the underlying mechanisms of the RIS, the prediction of the Marwick model was compared with the RIS measurements. The model with parametric input in accordance with the Inverse Kirkendall effect (IKE), i.e. a preferential flux of elements associated with vacancy flux, was inadequate in explaining the observed amount of segregation of Fe, Cr and Ni. Specifically, a considerably lower ratio of the concentration gradient between Cr and Fe, ∇Cr∇Fe, of 0.53 was obtained from the RIS measurements versus the model-predicted value of 2.3, indicating an overestimation of Cr depletion at the bubbles by the Marwick model. To explain this difference between the model prediction and the RIS measurements, in addition to the Cr depletion at the bubbles by IKE due to the fastest diffusion via vacancy flux, a preferential interstitial flux of Cr from the matrix to the bubbles is discussed as a possible concurrently operating mechanism.