Irradiated Steel Research Articles

Simulation of stainless ferritic-martensitic and austenitic steel hardening after irradiation in ion accelerator. Part 2. Development of a methodology for determining the ion mode irradiation of austenitic steels

A methodology for determining the irradiation mode for ferritic-martensitic steels at ion accelerator has been developed and experimentally substantiated, providing radiation hardening of these steels, identical to that realized under neutron irradiation. The change in Vickers microhardness is used as a measure of radiation hardening. The paper presents the results of a study of radiation-induced changes in the microhardness of austenitic steels 08Kh18Н10Т and 08Kh16Н20М2Т irradiated in reactors SM-3, VVER-440, BOR-60, SM-3 + BOR-60 to damaging doses of 10.2–33.7 dpa in the interval of temperatures 60–500°C. A study of radiation-induced changes in microhardness in a wider range of irradiation temperatures, post-irradiation annealing of irradiated steels was carried out in the range from 400 to 600°C, simulating irradiation at temperatures equal to annealing temperatures. Data are presented on radiation-induced changes in microhardness after irradiation in the ion accelerator of the State Scientific Center of the Russian Federation – Institute for Physics and Power Engineering named after A.I. Leypunsky (IPPE) with Ni+4 ions and He+ ions up to concentrations of 0–7 appm/dpa at damaging doses of 13–30 dpa and temperatures of 300–650°. A transition function has been established that connects the irradiation of temperatures during neutron and ion irradiation at a given damaging dose, ensuring the same radiation hardening of austenitic steels.

Simulation of stainless ferritic-martensitic and austenitic steel hardening after irradiation in ion accelerator. Part 1. Development of a methodology for determining the ion mode irradiation of ferritic-martensitic steels

A methodology for determining the irradiation mode for ferritic-martensitic steels at ion accelerator has been developed and experimentally substantiated, providing radiation hardening of these steels, identical to that realized under neutron irradiation. The change in Vickers microhardness is used as a measure of radiation hardening. A study was carried out of radiation-induced changes in the microhardness of ferritic-martensitic steels 07Kh12NMFB and EP-823 after neutron and ion irradiation to damaging doses of 10–30 dpa in the temperature range 350–600°C. These materials were irradiated with neutrons in the reactors BOR-60, BN-600 and in the ion accelerator of the State Scientific Center of the Russian Federation – Institute for Physics and Power Engineering named after A.I. Leypunsky (IPPE) with Fe3+, Fe4+ ions and He+ ions to concentrations of 0.2 and 4 appm/dpa. A transition function has been established that connects the irradiation temperatures for neutron and ion irradiation at a given damaging dose, ensuring the same radiation hardening.

On the theory of the nucleation of gas bubbles at grain boundaries and incoherent inclusions

On the base of the critical analysis of two-dimensional models of the nucleation of gas filled bubbles at grain boundaries of helium-implanted specimens under the action of tensile stresses, a new model is developed within the framework of the Reiss theory of homogeneous nucleation in binary systems. This approach considers that gas bubbles are formed as a result of agglomeration in a binary system of vacancies and gas atoms at grain boundaries, avoiding significant simplifications of previous models based on the classical nucleation theory for single-component (unary) systems. The new model is extended to consider the nucleation of Xe bubbles at grain boundaries in UO2 under irradiation conditions and can be used for numerical analysis of experimental observations after the foreseen implementation in a fuel performance code. A similar approach can be applied to the nucleation and growth of gas bubbles on incoherent inclusions, such as those observed in irradiated ODS steels.

Prediction of dK/da effects on stress corrosion crack growth rate of irradiated stainless steels based on slip oxidation mechanism

This study developed a predicting method for the crack growth rate (CGR) of stress corrosion cracking (SCC) under varying stress intensity factor (K) conditions for irradiated stainless steel (SS). First, optimization of the Hashimoto-Koshiishi model equation for calculating CGR under varying dK/da conditions was carried out using the Rice, Dugan, and Sham (RDS) equation for the crack tip strain rate. Second, the model input coefficients were set to incorporate the effect of dK/da on SCC CGR to fit the experimental data. Finally, the effect of dK/da on the CGR for 3 dpa irradiated SS of a core shroud was evaluated taking into account the weld residual stress distribution. The model prediction showed that there was little effect on the CGR for the increasing and decreasing K regions in the plate thickness direction of the core shroud under boiling water reactor (BWR) normal water chemistry, but the effect was not negligible for the decreasing K region under BWR hydrogen water chemistry.

How precisely are solute clusters in RPV steels characterized by atom probe experiments?

Atom probe tomography (APT) is a powerful microscopy technique to characterize nano-sized clusters of the alloying elements in the bulk of reactor pressure vessel (RPV) steels. These clusters are known to dominantly determine the evolution of mechanical properties under irradiation. The results are conventionally summarized as the overall number density N and the average diameter D of the solute clusters identified in the material. Here, we illustrate that these descriptors are intrinsically imprecise because they are steered by the parameters involved in the measurement and data processing, some of which are directly under the control of the operators, but some others not. Consequently, a direct comparison between data derived at different laboratories is compromised, and key trends such as the evolution with dose, are masked. This study relies on a state-of-the-art physical model for neutron irradiation in steels to make reliable estimates of the microstructure before the measurement is performed, which allows the prediction of the population of solute clusters that are not seen by APT. We mimic APT measurements from simulated microstructures, performing a detailed study of the effects of the parameters of the analysis. We show that the values of N and D reported in the scientific literature can be matched by the predictions of our theoretical model only if specific sets of parameters are used for each laboratory that issued the measurements. We also show that if, on the contrary, all studied cases are analyzed in a consistent way, the scatter of N and D values is reduced. Specifically, we find that the average diameter D is nearly a constant value with dose, independently of the material's chemical compositions, while N increases with dose, but is also influenced by other variables. The approach we developed and used proves to have added value as complement to APT experiments in reactor pressure vessel steels.

Identification of nano-sized precipitates in CLAM steel induced by 3.5 MeV Fe13+ ion irradiation

Nano-sized precipitate phases in normalized-and-tempered CLAM steel (HEAT-0912) before and after 3.5 MeV Fe13+ ion irradiation at 400 °C to 0.61, 1.29 and 2.77 dpa were studied using transmission electron microscope and high-resolution transmission electron microscope. Six types of nanoprecipitate phases based on Cr23C6, TaC, Ta3C2, VC, Cr7C3, and V2C were identified in unirradiated steel. During irradiation, a large number of nanoprecipitates uniformly distributed in the matrix occurred, and the number density of nanoprecipitates initially increased and subsequently decreased with increasing the irradiation dose. Nanoscale Cr-rich M23C6, V-rich MX and M2X phases but no nanoscale phases based on TaC, Ta3C2 and Cr7C3 were identified in the steel irradiated up to 2.77 dpa. After irradiation, five types of irradiation-induced nanoscale new phases were identified in the irradiated steel. These include Fe-rich M3X2 (Cr3C2 types I and II) carbonitrides with the same simple orthorhombic lattice and totally different lattice parameters (approximately a/b/c = 1.146/0.552/0.2821 nm and 0.285/0.925/0.696 nm), Fe-rich M2C, M3C and M5C2 carbides, but the occurrence of which under irradiation varied with the irradiation dose. Fe-rich M3X2 (Cr3C2 type I) and Fe-rich M5C2 phases were identified in the steel irradiated up to 1.29 dpa and only to 2.77 dpa, respectively. Fe-rich M2C and M3C in addition to Fe-rich M3X2 (Cr3C2 type II) phases were identified in the steel irradiated up to 2.77 dpa. The irradiation-induced Fe-rich M3X2 (Cr3C2 type I & II), M5C2 and M2C precipitates appear to be identified, for the first time, in reduced activation ferritic/martensitic steels including CLAM steel. The formation mechanism of these irradiation-induced precipitate phases is also discussed.

Cluster dynamics study on nano damage of RPV steels under proton irradiation at 290°C

Irradiation-induced defects such as dislocation loops, cavities or solute clusters and chemical composition segregation of reactor pressure vessel (RPV) steel are the root causes of irradiation embrittlement. Combining two nucleation mechanisms, namely, the uniform nucleation and non-uniform nucleation of solute clusters (such as Cu-rich phase), a cluster kinetic simulation was established based on the reaction rate theory, and the co-evolution of matrix damage and Cu-rich phase in low-copper RPV steel was simulated under irradiation. And the average size and number density of defective clusters and solute clusters were established with irradiation dose. Compared with the average size and number density of dislocation loops observed by transmission electron microscopy (TEM) of proton irradiated RPV steel at 290°C, the verification results show that the cluster dynamics model considering both the nucleation mechanism of interstitial dislocation loops and vacancy clusters can well simulate the irradiation damage behavior of materials.

Effects of interstitial cluster mobility on dislocation loops evolution under irradiation of austenitic steel

Effects of interstitial cluster mobility on dislocation loops evolution under irradiation of austenitic steel

Microstructural analysis of ex-service neutron irradiated stainless steel nuclear fuel cladding by high-speed AFM

Fuel cladding in advanced gas-cooled nuclear reactors is made of an austenitic stainless steel referred to as 20Cr/25NiNb. This material can undergo microstructural changes in these extreme environments through a combination of thermal effects and radiation damage. In this work, the microstructures present in an ex-service irradiated 20Cr/25Ni-Nb sample are studied using correlated high-speed atomic force microscopy (HS-AFM) and electron microscopy techniques for the first time. Correlated topographic, crystallographic, and chemical information from the sample surface enabled identification of secondary phase precipitates (SPPs) including M23C6, sigma phase, NbC, and G phase. These SPPs can have adverse effects on the material properties. HS-AFM analysis showed surface textures unique to each SPP. Voids formed due to irradiation were also identified across the surface by HS-AFM. These voids were found to be larger in size and depth along grain boundaries. Further analysis identified a relationship between void size and frequency and grain boundary misorientation and SPP presence.

Experimental and modeling study on irradiation effect of A508-Ⅲ steel

Irradiation can induce microscopic defects in reactor pressure vessel (RPV) steel, resulting in macroscopic hardening and embrittlement. This paper proposes a micro‑meso-macroscale calculation framework for the mechanical properties of irradiated RPV steel based on experimental and numerical investigations. The mean size and number density of Fe-ion-irradiation-induced dislocation loops were measured using transmission electron microscopy (TEM). These parameters were described by a power-function relationship with the irradiation dose at various temperatures and were used as input parameters for the crystal plasticity finite element model (CPFEM). Nano-indentation tests of A508-Ⅲ steel under different irradiation conditions were successfully simulated. In addition, microstructural characterization and molecular dynamics (MD) simulations revealed the phenomenon of dislocation pinning and the sweeping of irradiation defects during dislocation migration, implying that the dislocation-loop interactions could be responsible for the formation of dislocation channels and curved dislocation lines. Furthermore, increase in yield stress and the ductile-brittle transition temperature (DBTT) at various temperatures were investigated. Results indicated that yield stress, DBTT, and the degree of irradiation hardening and embrittlement increased with the irradiation dose and decreased with irradiation temperature. Finally, the adjusted reference temperature (ART), a significant design parameter for nuclear reactors, was calculated and evaluated using the multiscale framework. The framework can serve as a reference text for extending the life of nuclear reactors.

Irradiation Effects on Tensile Properties of Reduced Activation Ferritic/Martensitic Steel: A Micromechanical-Damage-Model-Based Numerical Investigation

The tensile properties of reduced activation ferritic/martensitic (RAFM) steels are significantly influenced by neutron irradiation. Here, a mechanism-based model taking account of the typical ductile damage process of void nucleation, growth, and coalescence was used to study the temperature and irradiation effects. The elastic–plastic response of RAFM steels irradiated up to 20 dpa was investigated by applying the GTN model coupled with different work hardening models. Through a numerical study of tensile curves, the GTN parameters were identified reasonably and satisfying simulation results were obtained. A combination of Swift law and Voce law was used to define the flow behavior of irradiated RAFM steels. The deformation localization could be adjusted effectively via setting the nucleation parameter εn close to the strain where necking occurs. Because εn changed with uniform elongation, εn decreased with the testing temperature and rose with an irradiation temperature above 300 °C. The nucleation parameter fn increased with the testing temperature for RAFM steels before irradiation. For irradiated RAFM steels, fn barely changed when the irradiation temperature was below 300 °C and then it rose at a higher irradiation temperature. Meanwhile, the ultimate strength of the simulated and experimental curves showed good agreement, indicating that this method can be applied to engineering design.

The effect of hydrogen-helium synergism on dislocation loops in RAFM steel at different temperatures by dual-beam ion irradiation

Simultaneous Fe+H ion irradiation (denoted as Fe+H) and Fe+H ion irradiation with pre-implantation of He (denoted as He/Fe+H) were carried out on Chinese Low-Activation Martensitic (CLAM) steels at 350 ℃-550 ℃ to investigate the synergistic effect between H and He. The peak dose was 15 dpa, the He concentration in the observation area was 11 appm/dpa and the H concentration was 44 appm/dpa. The microstructure and hardening of the irradiated samples were evaluated by TEM and nanoindentation, respectively. Bubbles were not observed in irradiated CLAM steels at all temperatures, which implies that the promotion of bubbles by H may not be as strong as that by He and that H attenuates the promotion of bubbles by He. The hardening induced by He/Fe+H irradiation at 450 °C is about 1.6 times that of Fe+H irradiation, the synergistic effect of H and He is most significant at this temperature. Comparison with Fe+He irradiation and single Fe irradiation reveals that Fe+H irradiation at 350 °C induces lower interstitial defect damage than Fe+He irradiation, but the level of interstitial defects induced by the synergistic effect of both H and He with displacement damage is comparable at higher temperatures. The average size of dislocation loops induced by Fe+H irradiation is between that of single Fe irradiation and Fe+He irradiation, but the number density of dislocation loops induced by Fe+H irradiation is higher than that of Fe+He irradiation. The synergistic effect of He with displacement damage tends to promote the growth of dislocation loops, while the synergistic effect of H tends to promote the nucleation of dislocation loops.

Refinement MRP-135 constitutive equation model for tensile properties of austenitic stainless steel

This study focuses on analyzing and enhancing MRP-135 constitutive models, which are crucial for assessing the tensile properties of irradiated stainless steels in the reactor vessel internals of pressurized water reactors. The newly developed model, named M1, incorporates parameters from MRP-135, while minimizing modifications to ensure high compatibility with existing models. A significant improvement in M1 is the resolution of the inversion phenomenon observed between the yield strength and ultimate tensile strength at low temperatures in Type 316 stainless steel, a challenge previously identified in MRP-135 Rev. 2. M1 integrates the temperature effects from the earlier version, MRP-135 Rev. 1, to address this issue effectively. Additionally, for Type 316, M1 refined the equations to represent the cold work factor more accurately, enabling precise interpolation of strength variations across different levels of cold work from solution-annealed to cold-worked conditions. A comparative analysis of the M1 and the MRP-135 models against actual irradiated tensile test data for Type 304 and 316 stainless steel demonstrated the superior accuracy of M1, as indicated by a lower root-mean-square deviation. This study highlights the importance of continuous improvements in constitutive models, with future work aimed at further refining these models based on comprehensive irradiation test data.

Size dependence of micro-scale mechanical properties on heavy-ion irradiated tempered-martensitic steel evaluated through nanoindentation and micropillar compression tests

The size dependence of microscale mechanical properties is crucial for evaluating irradiation hardening using nanoindentation and micropillar compression tests. To predict bulk hardness without the indentation size effect in heavy-ion irradiated materials, this study designed a modified Nix–Gao model incorporating the effect of irradiation hardening. The model was used to determine the bulk hardness of irradiated film and unirradiated substrate in tempered F82H martensitic steel irradiated with 10.5 MeV Fe-ion. The bulk hardness of unirradiated F82H predicted from the nanoindentation hardness was consistent with the traditional Vickers hardness obtained using a multiscale hardness relationship. The effects of irradiation conditions and the irradiation damage gradient were also determined by the modified model. In a micropillar compression test performed using micropillars with sizes of 1–8 μm, the flow stress at 8 % plastic strain of the micropillars showed a stable and comparable value with that of a bulk tensile specimen, while the yield stress at 0.2 % strain offset decreased with decreasing pillar sizes. The irradiation hardening calculated using the bulk hardness obtained in the nanoindentation tests and the flow stress determined from the micropillar compression tests at 8 % plastic strain were consistent, as predicted by Tabor's law. These findings prove that the proposed method is effective for assessing the mechanical characteristics of a size-limited region, even on the subsurface of metallic materials exposed to heavy-ion irradiation.

Impact of trace silicon on irradiation hardening and embrittlement of RAFM steel subjected to neutron irradiation

The content of silicon (Si) element in reduced activation ferritic/martensitic (RAFM) steels varies significantly among different candidates of fusion reactor structural materials. Si is not intentionally added to the RAFM steel but rather remains as a residue during the deoxidation process in steel fabrication. To control and reduce the Si content, removing processes and special treatments are required. The additional refining processes significantly increase the cost of steel fabrication. Currently, the role of residual or trace Si in neutron irradiation of steel remains unclear. In the presented study, two 9Cr-2 W RAFM steels with Si contents of 0.1 wt.% and 0.01 wt.% were fabricated and subjected to a neutron irradiation at 563 K with a neutron fluence of 4.8 × 1023m−2. The main results show that 0.1 wt.% Si can contribute to the suppression of irradiation hardening and embrittlement, as evidenced by the reduction in irradiation-increased hardness and yield strength, while maintaining preferable elongation. Microscopic analysis suggests that an appropriate content of Si helps refine the grains in RAFM steel, increases the ductile fracture region of the fracture surface, and effectively reduces the density of dislocation loops induced by neutron irradiation. The findings evidence that retaining a certain amount of trace Si should be an informed and rational choice in the design of RAFM steel.

Evaluation of the Embrittlement in Reactor Pressure-Vessel Steels Using a Hybrid Nondestructive Electromagnetic Testing and Evaluation Approach.

The nondestructive determination of the neutron-irradiation-induced embrittlement of nuclear reactor pressure-vessel steel is a very important and recent problem. Within the scope of the so-called NOMAD project funded by the Euratom research and training program, novel nondestructive electromagnetic testing and evaluation (NDE) methods were applied to the inspection of irradiated reactor pressure-vessel steel. In this review, the most important results of this project are summarized. Different methods were used and compared with each other. The measurement results were compared with the destructively determined ductile-to-brittle transition temperature (DBTT) values. Three magnetic methods, 3MA (micromagnetic, multiparameter, microstructure and stress analysis), MAT (magnetic adaptive testing), and Barkhausen noise technique (MBN), were found to be the most promising techniques. The results of these methods were in good agreement with each other. A good correlation was found between the magnetic parameters and the DBTT values. The basic idea of the NOMAD project is to use a multi-method/multi-parameter approach and to focus on the synergies that allow us to recognize the side effects, therefore suppressing them at the same time. Different types of machine-learning (ML) algorithms were tested in a competitive manner, and their performances were evaluated. The important outcome of the ML technique is that not only one but several different ML techniques could reach the required precision and reliability, i.e., keeping the DBTT prediction error lower than a ±25 °C threshold, which was previously not possible for any of the NDE methods as single entities. A calibration/training procedure was carried out on the merged outcome of the testing methods with excellent results to predict the transition temperature, yield strength, and mechanical hardness for all investigated materials. Our results, achieved within the NOMAD project, can be useful for the future potential introduction of this (and, in general, any) nondestructive evolution method.

Study of stress corrosion cracking susceptibility of irradiated ferrite-martensitic stainless steel 07Kh12NMFB in supercritical water. Part 2. Development of corrosion crack identification technique and analysis of autoclave test results

The studies of stress corrosion cracking have been carried out for the stainless ferritic-martensitic steel with chromium content of 12% irradiated to a damage neutron dose of ~12 dpa. This steel was chosen as a candidate material for the internals of supercritical water-cooled reactors (SWCR).The first part of the paper was devoted to autoclave testing of specially designed disk specimens under constant load in a supercritical water environment (at 450°C and 250 atm pressure). In this part of the article the developed technique for corrosion cracks identification is presented and analysis of autoclave tests results is performed.As a result of the experiments performed, the threshold stress values below which stress corrosion cracking initiation doesn’t occur. The values of threshold stresses are determined for the studied steel irradiated at temperatures of 390 and 550°C. Possible mechanisms of stress corrosion cracking of the studied steel are analyzed and directions for further research are proposed.

Study on fracture assessment by GTN model for BWR reactor internals based on material properties of simulated irradiated stainless steels

The fitness for service code of the Japan Society of Mechanical Engineers (JSME) provides a fracture evaluation methodology based on linear elastic fracture mechanics (LEFM) for boiling water reactor (BWR) internal components irradiated to a neutron fluence of 3.0 × 1024 n/m2 or more. On the other hand, it is known that austenitic stainless steel maintains its ductility and toughness even after neutron irradiation up to the fluence expected for BWR internal components. Accordingly, it is preferable to apply the ductile fracture evaluation methodology for BWR internal components. The present study examined the applicability of damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel. The changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. As a result, in terms of material strength properties the applicability of the GTN model was confirmed for irradiated stainless steel using the simulated material for neutron fluence up to about 1.5 × 1025 n/m2.

Determination of Mo-93 inventory in irradiated BWR tie plate using triple quadrupole ICP-MS

The inventory of radionuclides is a crucial input parameter for the safety assessment of irradiated metal waste. Among the radionuclides, Mo-93 is remarkable radionuclide owing to its long half-life (4,000 years) and poor sorptive property (anionic species). A method for determining Mo-93 by measuring characteristic X-rays was proposed previously; however, this method requires complex sequential and/or chromatographic separation of Mo-93. In the present study, triple quadrupole inductively coupled plasma mass spectrometry (ICP-MS/MS)—a leading-edge method—was used to determine the inventory of Mo-93 in an irradiated boiling water reactor (BWR) tie plate (stainless steel) for the first time. Furthermore, an activation calculation for Mo-93 was conducted using ORIGEN2. The results indicated that the inventory of Mo-93 in the BWR tie plate was 690 ± 83 Bq/g. This study demonstrates that triple quadrupole ICP-MS allows the simple quantitative analysis of Mo-93 in irradiated stainless steel without complex chemical separation.Graphical abstract

Development of pressure-temperature limit curves considering unified constraint for reactor pressure vessel

In this work, the pressure-temperature (P-T) limit curves considering unified constraint for a typical large-sized reactor pressure vessel (RPV) with axial surface cracks have been investigated and developed based on the KJ-Ad* two-parameter and the Master Curve (MC) method. The ASME fracture toughness KIc curve and the MC with and without considering unified constraint for irradiated RPV steel were used in the constructions of the P-T limit curves. The effects of crack sizes (constraint levels) and fracture toughness curves on the P-T limit curves have been analyzed for the RPV under heatup and cooldown conditions. The results further show that the P-T limit curves based on traditional ASME KIc curve has the highest degree of conservatism and the use of the MC method can significantly increase the safety margin of P-T limit curves. With increasing crack sizes (increasing constraint levels), the safety region in P-T limit curves reduces, which may lead to the decrease of operating flexibility for nuclear power plants. The accuracy of the P-T limit curves can be improved by considering unified constraint. For smaller size cracks with lower constraint in the RPV, the consideration of constraint can reduce conservatism of the P-T limit curves, and for larger size cracks with higher constraint, it can avoid the non-conservatism. It is recommended to obtain more accurate P-T limit curves based on the MC with considering unified constraint and finite element analysis for RPVs under various operating conditions.