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Reactor pressure vessel (RPV) steels are highly susceptible to irradiation embrittlement due to prolonged exposure to high temperature, high pressure, and intense neutron irradiation. This leads to the shift in nil-ductility transition reference temperature—∆RTNDT. The change in ∆RTNDT follows a certain distribution pattern and is impacted by factors including chemical composition, neutron fluence, and irradiation temperature. Existing empirical procedures can estimate ∆RTNDT based on fitting extensive irradiation embrittlement data, but their reliability has not been thoroughly investigated. Probability statistical distributions and the Gamma stochastic process were performed to model material property degradation in RPV steels from a pressurized water reactor due to irradiation embrittlement, with the probability models considered being normal, Weibull, and lognormal distributions. Comparisons with existing empirical procedures showed that the Weibull distribution model and the Gamma stochastic model demonstrate good reliability in predicting ∆RTNDT for RPV steels. This provides a valuable reference for studying irradiation embrittlement in RPV materials. [ABSTRACT FROM AUTHOR]

A sophisticated Bayesian multilevel model for estimating group bias was developed to improve the utility of the ASTM E900-15 embrittlement trend curve (ETC) to assess the conditions of nuclear power plants (NPPs). For multilevel model development, the Baseline 22 surveillance dataset was basically classified into groups based on the NPP name, product form, and notch orientation. By including the notch direction in the grouping criteria, the developed model could account for TTS differences among NPP groups with different notch orientations, which have not been considered in previous ETCs. The parameters of the multilevel model and biases of the NPP groups were calculated using the Markov Chain Monte Carlo method. As the number of data points within a group increased, the group bias approached the mean residual, resulting in reduced credible intervals of the mean, and vice versa. Even when the number of surveillance test data points was less than three, the multilevel model could estimate appropriate biases without overfitting. The model also allowed for a quantitative estimate of the changes in the bias and prediction interval that occurred as a result of adding more surveillance test data. The biases estimated through the multilevel model significantly improved the performance of E900-15.

Reactor pressure vessel (RPV) steels are highly susceptible to irradiation embrittlement due to prolonged exposure to high temperature, high pressure, and intense neutron irradiation. This leads to the shift in nil-ductility transition reference temperature—∆RTNDT. The change in ∆RTNDT follows a certain distribution pattern and is impacted by factors including chemical composition, neutron fluence, and irradiation temperature. Existing empirical procedures can estimate ∆RTNDT based on fitting extensive irradiation embrittlement data, but their reliability has not been thoroughly investigated. Probability statistical distributions and the Gamma stochastic process were performed to model material property degradation in RPV steels from a pressurized water reactor due to irradiation embrittlement, with the probability models considered being normal, Weibull, and lognormal distributions. Comparisons with existing empirical procedures showed that the Weibull distribution model and the Gamma stochastic model demonstrate good reliability in predicting ∆RTNDT for RPV steels. This provides a valuable reference for studying irradiation embrittlement in RPV materials.

Most WWER-440 reactors are operated over the planned operational lifetime of 30 years. Now, the owners/operators are preparing for the second phase of extending the operation. The paper presents an overview of the most important aspects of ageing of the main components of the WWER-440-type reactors based on the operational and ageing management experiences and primary research efforts supporting the operation. The paper aims to demonstrate that the expectations for the operability of these reactors for up to 60–70 years are realistic. The scope of the review is limited to the ageing of the reactor pressure vessel, vessel internals, and steam generator, which are the lifetime-limiting components besides the containment structure. Some of the corrective actions, improvements of in-service inspections, maintenance, and ageing management are also considered in the paper. It is shown that the selection of materials and unavoidable material degradation processes do not limit the operation of the WWER-440 reactors for up to 60–70 years under the established operational environment and ageing management practice. [ABSTRACT FROM AUTHOR]

LT21 a type of aluminum alloy used for the irradiation channel of the first heavy water research reactor (HWRR) in China. Studying the mechanical property of irradiated LT21 aluminum under actual service conditions is essential for evaluating its application property. In this paper, tensile specimens of irradiated LT21 were manufactured from the decommissioned irradiation channel of an HWRR; then, tensile tests were carried out, and then the fracture surfaces were observed. The effect of neutron irradiation on tensile behavior and the failure mechanism was analyzed by comparing the result of irradiated and unirradiated LT21 specimens. The results show that, with the thermal neutron flux increasing to 2.38 × 1022 n/cm2, the YS gradually increased from the initial 158 MPa to 251 MPa, the UTS increased from 262 MPa to 321 MPa, and the elongation decreased from 28.8% to about 14.3%; the brittle fracture of the LT21 specimen appeared after irradiation, and the proportion of brittle fracture increased as the neutron fluence increased; the nanophase structures, with a size of less than 50 nm, were precipitated in the LT21 aluminum alloy after neutron irradiation. Transmutation Si is presumed to be the main cause of the radiation effect mechanism of LT21. [ABSTRACT FROM AUTHOR]

The development of prediction model on irradiation embitterment (PMIE) of reactor pressure vessel (RPV) is an important method for nuclear reactor long term operation. Based on the physical mechanism of RPV irradiation embrittlement, a preliminary model is determined and the critical threshold of Cu content of 0.072% is obtained according to this preliminary model. Then a prediction model named PMIE-2020 for low Cu RPV steels is developed. At last the residual, standard deviation and predicted values and test values distribution analysis are given. Simultaneously, a comparison between PMIE-2020 and other prediction model and irradiation data is provided. Results indicate that the predicted results of PMIE-2020 has no tendency with influence factors such as neutron fluence, flux, irradiation temperature, chemical elements Cu, P, Mn, Ni, Si. The residual standard deviation is 10.76 °C, which is lower than present prediction model. The distribution between predicted values of PMIE-2020 and test values are located the area near the 45° line. These results prove that the PMIE-2020 have high accuracy on irradiation embrittlement prediction.

In the structural integrity assessment of an embrittled reactor pressure vessel (RPV) based on probabilistic fracture mechanics (PFM), it is crucial to consider both of the mean value and the uncertainty of the embrittlement prediction. Typically, the embrittlement prediction uncertainty is given by a normal distribution, and its standard deviation is determined from the residuals of the measured and predicted values of all data used to develop the embrittlement prediction method. Therefore, the same uncertainty is assumed regardless of neutron fluence and the available data. To overcome this issue, the Japan Atomic Energy Agency recently developed the embrittlement prediction method based on machine learning and Bayesian statistics, i.e., the Bayesian nonparametric (BNP) method, and introduced it to the PFM analysis code. The BNP method can predict the probability distribution according to data scarcity and provide more reasonable uncertainty because it estimates significant uncertainties where the scatter of actual measured data is large and the number of data is small. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor is conducted using the PFM analysis of structural components in aging light water reactors, where the embrittlement probability distribution calculated using the BNP method is used. Furthermore, the effects of different embrittlement prediction methods and uncertainties on the failure frequency of RPVs are investigated in the PFM analyses, and the results are presented. • Embrittlement prediction based on machine learning was applied to probabilistic fracture mechanics analysis. • The probability distribution of superposition of normal distributions according to the number and variation of the data was used. • The effect of different embrittlement prediction methods and uncertainties on the failure frequency of RPVs was investigated. [ABSTRACT FROM AUTHOR]

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy's unique balance of high fracture toughness and high tungsten content; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. To alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni+ and He+ ion irradiations to mimic the expected displacement damage and He gas production expected after five years of service as materials for plasma facing components. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and a hexagonal W 2 C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation enhanced and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material's overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases. [Display omitted] [ABSTRACT FROM AUTHOR]

In this study, machine learning (ML) techniques were used to model surveillance test data of nuclear power plants from an international database of the ASTM E10.02 committee. Regression modeling was conducted using various techniques, including Cubist, XGBoost, and a support vector machine. The root mean square deviation of each ML model for the baseline dataset was less than that of the ASTM E900-15 nonlinear regression model. With respect to the interpolation, the ML methods provided excellent predictions with relatively few computations when applied to the given data range. The effect of the explanatory variables on the transition temperature shift (TTS) for the ML methods was analyzed, and the trends were slightly different from those for the ASTM E900-15 model. ML methods showed some weakness in the extrapolation of the fluence in comparison to the ASTM E900-15, while the Cubist method achieved an extrapolation to a certain extent. To achieve a more reliable prediction of the TTS, it was confirmed that advanced techniques should be considered for extrapolation when applying ML modeling.

The prediction of irradiation-induced transition temperature shift for RPV steels is an important method for long term operation of nuclear power plant. Based on the irradiation embrittlement data, an irradiation-induced transition temperature shift prediction model is developed with machine learning method XGBoost. Then the residual, standard deviation and predicted value vs. measured value analysis are conducted to analyze the accuracy of this model. At last, Cu content threshold and saturation values analysis, temperature dependence, Ni/Cu dependence and flux effect are given to verify the reliability. Those results show that the prediction model developed with XGBoost has high accuracy for predicting the irradiation embrittlement trend of RPV steel. The prediction results are consistent with the current understanding of RPV embrittlement mechanism.

Most WWER-440 reactors are operated over the planned operational lifetime of 30 years. Now, the owners/operators are preparing for the second phase of extending the operation. The paper presents an overview of the most important aspects of ageing of the main components of the WWER-440-type reactors based on the operational and ageing management experiences and primary research efforts supporting the operation. The paper aims to demonstrate that the expectations for the operability of these reactors for up to 60–70 years are realistic. The scope of the review is limited to the ageing of the reactor pressure vessel, vessel internals, and steam generator, which are the lifetime-limiting components besides the containment structure. Some of the corrective actions, improvements of in-service inspections, maintenance, and ageing management are also considered in the paper. It is shown that the selection of materials and unavoidable material degradation processes do not limit the operation of the WWER-440 reactors for up to 60–70 years under the established operational environment and ageing management practice.

The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works. [ABSTRACT FROM AUTHOR]

Existing elastic analysis methods cannot be adhered to in order to assess the structural integrity of a reactor vessel and internals for a beyond design basis earthquake. Elasto-plastic analysis methods are required, and the factors that affect the elasto-plastic behavior of reactor materials should be taken into account. In this study, a material behavior model was developed that considers the irradiation embrittlement effect, which affects the elasto-plastic behavior of the reactor material. This was used to perform the elasto-plastic time history analyses of the reactor vessel and its internals for beyond design basis earthquake. For this investigation, appropriate beyond design basis earthquakes and reliable finite element models were used. Based on the analysis results, consideration was given to the load reduction effect and the margin change. These were transferred to the internals due to the plastic deformation of the reactor vessel.

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. • Damage mechanics analysis based on the Gurson‒Tvergaard‒Needleman (GTN) model for irradiated stainless steel was investigated. • Changes of mechanical properties of the material due to neutron irradiation were simulated by cold working. • Fracture tests and analyses were performed with plate specimens. • Applicability of GTN model was examined by comparing the maximum loads and fracture behaviors between the tests and analyses. • 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. [ABSTRACT FROM AUTHOR]

LT21 a type of aluminum alloy used for the irradiation channel of the first heavy water research reactor (HWRR) in China. Studying the mechanical property of irradiated LT21 aluminum under actual service conditions is essential for evaluating its application property. In this paper, tensile specimens of irradiated LT21 were manufactured from the decommissioned irradiation channel of an HWRR; then, tensile tests were carried out, and then the fracture surfaces were observed. The effect of neutron irradiation on tensile behavior and the failure mechanism was analyzed by comparing the result of irradiated and unirradiated LT21 specimens. The results show that, with the thermal neutron flux increasing to 2.38 × 1022 n/cm2, the YS gradually increased from the initial 158 MPa to 251 MPa, the UTS increased from 262 MPa to 321 MPa, and the elongation decreased from 28.8% to about 14.3%; the brittle fracture of the LT21 specimen appeared after irradiation, and the proportion of brittle fracture increased as the neutron fluence increased; the nanophase structures, with a size of less than 50 nm, were precipitated in the LT21 aluminum alloy after neutron irradiation. Transmutation Si is presumed to be the main cause of the radiation effect mechanism of LT21.

Reduced-pressure electron beam (EB) plate butt welds were manufactured in two low-alloy pressure-vessel steels, SA508 Gr 3 Cl 1 and SA508 Gr 2, at two thicknesses in both steels, 30 mm and 130 mm. Transient temperatures during welding were recorded using thermocouple arrays. Residual stresses in the as-welded condition and after post-weld heat treatment were measured using diverse methods: neutron diffraction and the contour method at 30 mm thickness; and deep hole drilling and the contour method at 130 mm. Incremental centre hole drilling measurements were performed at 130 mm thickness to better understand near-surface stresses. Weld and heat-affected zone microstructures and microconstituents were evaluated using a combination of hardness mapping, optical microscopy and electron microscopy. The as-welded residual stresses exhibit the characteristic M-shaped distribution for hardenable steels, reaching 500–600 MPa in tension in both steels at both thicknesses. However, the modest changes to the chemical composition and the change in plate thickness both significantly influenced microstructures, mechanical properties and residual stress distributions. These sensitivities underline the need for physically faithful models. This extensive characterisation study enables the development and validation of models that predict the development of microstructures and residual stresses in EB welds in low alloy pressure vessel steels.

• The changes in fracture behavior of stainless steel with in neutron dose can be predicted from the presented method. • The multiaxial fracture strain model is determined by simulating tensile and fracture toughness test. • The neutron irradiation embrittlement constant is determined by simulating tensile test. • The neutron irradiation embrittlement constant is multiplied with the multiaxial fracture strain model to simulate fracture toughness tests, enabling the prediction of degradation in fracture toughness due to neutron irradiation. This paper proposes a numerical method for predicting the reduction in the fracture toughness with neutron dose using the concept of an embrittlement constant. Tensile and fracture toughness test data for type 304 stainless steel in the literature [ 41 , 42 ] are used to validate the proposed method. Fracture toughness tests are simulated using a stress-modified fracture strain damage model implemented in an ABAQUS user-subroutine. The true stress-strain curve and one point of the fracture strain locus are obtained from the tensile test simulation. The remaining parameters of the fracture strain locus are determined by simulating a fracture toughness test. The fracture strain locus is modified to reflect the irradiation effect on the fracture toughness using the proposed irradiation embrittlement constant C I. The function of the constant with the neutron dose is calculated from the tensile tests and simulations. Subsequently, the modified damage model is used to predict the reduction in the fracture toughness. The predicted simulation results are compared with experimental data, showing good agreement. [Display omitted] [ABSTRACT FROM AUTHOR]

The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works.

The comparison of experimental values of the critical brittle temperature ΔTF and reference temperature ΔT0 of VVER-1000 reactor vessel weld metal with an elevated content of manganese and nickel is performed. ΔTF and ΔT0 values are defined proceeding from the standard impact bend Charpy and Charpy cracked fracture toughness specimen tests, respectively. Specimens were irradiated in industrial reactors in the frame of surveillance specimen program up to the fast (E ≥ 0.5 MeV) neutron fluences corresponding to the NPP long term operation period. The research results showed the shifts ΔTF and ΔT0 to agree with each other. Besides, it was discovered that in the range of over-design fluences the design embrittlement model has a tendency to underestimate the critical brittle temperature shift. [ABSTRACT FROM AUTHOR]

The Japan Electric Association code, JEAC 4206-2016 specifies a procedure for assessment of a reactor pressure vessel integrity against pressurized thermal shock (PTS) events which postulates a semi-elliptical flaw (10 mm in depth x 60 mm in length) beneath the cladding. JEAC 4206-2016 prescribes a simple method for calculating stress intensity factor of the flaw, taking into account the effect of yielding of cladding which has been originally developed in France. However, since this method is a rather conservative in comparison with that obtained by elastic-plastic finite element analysis, JEAC 4206-2016 also allows to calculate the stress intensity factor by converting the J-integral obtained from elastic-plastic finite element analysis. Since the J-integral is only defined for a linear or nonlinear elastic behavior material, the J-integral cannot be determined for an elastic-plastic behavior material during the unloading phase of PTS events. In addition, the residual stress and plastic strain induced by clad welding and joint welding needs to be considered in calculating the stress intensity factor. In this paper, the calculation procedure for a stress intensity factor of a fully circumferential flaw beneath the cladding during PTS events using finite element analysis is investigated.

The microstructure and nanomechanical properties of three samples of the modified stainless steel (referred as the alloy D9) were systematically characterized after neutron irradiation in the Advanced Test Reactor. The samples were irradiated to 5.0, 8.2, and 9.2 displacements per atom at 448, 430, and 683°C, respectively. The evolutions of dislocation loops, cavities, and radiation-induced precipitates were quantitatively studied to reveal their dose and temperature dependencies. Nanohardness and nanoindentation creep tests were conducted at room temperature on the irradiated samples. Unexpected radiation hardening was observed in the highest-temperature-irradiated sample due to the formation of an unknown type of Ni- and Si-rich precipitates whose contributions to the radiation and mechanical performances of the alloy were discussed. We provide the radiation-microstructure-property correlations of alloy D9 with new insights, which can benefit the development and optimization of advanced austenitic alloys for future nuclear applications. Image, graphical abstract [ABSTRACT FROM AUTHOR]

An improved Cluster Dynamics (CD) model of Cu rich and Mn–Ni–Si phase co-precipitation was developed to provide insights on the combined effects of the flux, fluence, temperature and alloy composition on irradiation enhanced precipitation leading to the embrittlement of reactor pressure vessel steels. The CD model was calibrated using a large microstructural database, and key parameters (e.g., interfacial energies) were fitted to minimize the predicted versus measured errors. The CD model was further validated against data not used in the fitting. The CD model predicts that: a) even 0.05% Cu reduces the Mn, Ni, Si precipitation threshold fluence; b) precipitate number densities increase, while their sizes and volume fractions, decrease with increasing flux; c) precipitate number densities and volume fractions increase with decreasing temperature; and, d) most of the matrix Cu precipitates in the early years of vessel service, while MnNiSi precipitates continue to grow up to very high extended life fluence. Image 1 [ABSTRACT FROM AUTHOR]

The reactor pressure vessel is made of ferritic low alloy steel, which has fracture behaviors that include ductile to brittle transition temperature and decrease of fracture toughness in the beltline region of the reactor pressure vessel steel due to irradiation embrittlement. Decrease of fracture toughness due to irradiation during plant operation needs to be taken into account in the assessment of the reactor pressure vessel against pressurized thermal shock events. A large number of fracture toughness data were accumulated in domestic PWR plant surveillance program tests and knowledge on the fracture toughness evaluations, such as scattering, were obtained through research works. In this study, fracture toughness curve for the assessment of domestic reactor pressure vessels were developed taking into consideration the latest knowledge on fracture toughness evaluations, including investigations on fracture toughness data scattering and fracture toughness curve reliability.

It has long been recognized that exposure to irradiation environments could dramatically degrade the mechanical properties of nuclear structural materials, i.e., irradiation-hardening and embrittlement. With the development of numerical simulation capability and advanced experimental equipment, the mysterious veil covering the fundamental mechanisms of irradiation-hardening and embrittlement has been gradually unveiled in recent years. This review intends to offer an overview of the fundamental mechanisms in this field at moderate irradiation conditions. After a general introduction of the phenomena of irradiation-hardening and embrittlement, the formation of irradiation-induced defects is discussed, covering the influence of both irradiation conditions and material properties. Then, the dislocation-defect interaction is addressed, which summarizes the interaction process and strength for various defect types and testing conditions. Moreover, the evolution mechanisms of defects and dislocations are focused on, involving the annihilation of irradiation defects, formation of defect-free channels, and generation of microvoids and cracks. Finally, this review closes with the current comprehension of irradiation-hardening and embrittlement, and aims to help design next-generation irradiation-resistant materials.

Reactor pressure vessel (RPV) is the critical un-changeable component of the reactor during its service lifetime, which prevents the radioactive leak of the nuclear power plant core. The irradiation test (about 10 × 10 19 cm − 2 , E > 1 MeV) in research reactor of the pressure vessel material was carried out, and the charpy impact test has been carried out before and after the neutron irradiation. Analysis of the impact energy and the fracture morphology has been done to estimate the embrittlement due to neutron irradiation. It shows that the main effects of neutron irradiation to fracture are the crack initiation and stable expansion process. And there also are a small amount of intergranular fracture in the unstable crack expansion after neutron irradiated which aware us pay more attention to the increasing intergranular fracture behavior of higher neutron fluence level after 60a nuclear power plant operation. [ABSTRACT FROM AUTHOR]

The development of prediction model on irradiation embitterment (PMIE) of reactor pressure vessel (RPV) is an important method for nuclear reactor long term operation. Based on the physical mechanism of RPV irradiation embrittlement, a preliminary model is determined and the critical threshold of Cu content of 0.072% is obtained according to this preliminary model. Then a prediction model named PMIE-2020 for low Cu RPV steels is developed. At last the residual, standard deviation and predicted values and test values distribution analysis are given. Simultaneously, a comparison between PMIE-2020 and other prediction model and irradiation data is provided. Results indicate that the predicted results of PMIE-2020 has no tendency with influence factors such as neutron fluence, flux, irradiation temperature, chemical elements Cu, P, Mn, Ni, Si. The residual standard deviation is 10.76 °C, which is lower than present prediction model. The distribution between predicted values of PMIE-2020 and test values are located the area near the 45° line. These results prove that the PMIE-2020 have high accuracy on irradiation embrittlement prediction., Competing Interests: The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (© 2023 The Authors.)

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

Formation of a high density of Mn–Ni–Si nanoscale precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels could lead to severe unexpected embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement prediction models, would emerge only at high fluence. However, the mechanisms and variables that control Mn–Ni–Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni contents were carried out at ∼295 °C to high and very high neutron fluences of ∼1.3 × 10 20 and ∼1.1 × 10 21 n cm −2 . Atom probe tomography shows that significant mole fractions of Mn–Ni–Si-dominated precipitates form in the Cu-bearing steels at ∼1.3 × 10 20 n cm −2 , while they are only beginning to develop in Cu-free steels. However, large mole fractions of these precipitates, far in excess of those found in previous studies, are observed at 1.1 × 10 21 n cm −2 at all Cu contents. At the highest fluence, the precipitate mole fractions primarily depend on the alloy Ni, rather than Cu, content. The Mn–Ni–Si precipitates lead to very large increases in measured hardness, corresponding to yield strength elevations of up to almost 700 MPa. [ABSTRACT FROM AUTHOR]

Highlights: [•] Ductile to brittle toughness predictions compared to experimental results. [•] Improved criterion for particle failure including measured particle distributions. [•] Predict experimentally measured locations of cleavage initiators. [•] Predicts toughness in transition region and shift with irradiation. [•] All results are obtained without changes in model parameters. [ABSTRACT FROM AUTHOR]

Abstract: We present a microstructural study of a surveillance test specimen from a reactor pressure vessel steel of a Russian-type nuclear reactor after neutron irradiation, post-irradiation annealing and re-irradiation, using atom probe tomography (APT) and positron annihilation spectroscopy (PAS). The APT results showed the formation of Cu-rich solute nano-clusters (CRCs) during the initial irradiation and their subsequent coarsening during annealing. After re-irradiation, new CRCs have been observed. The irradiation-hardening almost recovered during annealing. However, by re-irradiation, hardening comparable to that by the initial irradiation was observed. The hardening due to the CRCs formed during the initial irradiation, estimated using the Russell–Brown model, was almost the same as that observed. However, the estimated hardening after the re-irradiation was about half of the measured one. The other hardening is attributed to the newly formed MDs by the re-irradiation, which was evidenced by PAS. [Copyright &y& Elsevier]

[Display omitted] • Modest variations in SA508 steel chemistry have a marked impact on EB weld residual stresses. • Significant stresses remain after an ASME-compliant post-weld heat treatment. • Single-pass EB welds are likely to have thickness-dependent residual stress distributions. • Phase transformation kinetics must be incorporated into models for stress development. Reduced-pressure electron beam (EB) plate butt welds were manufactured in two low-alloy pressure-vessel steels, SA508 Gr 3 Cl 1 and SA508 Gr 2, at two thicknesses in both steels, 30 mm and 130 mm. Transient temperatures during welding were recorded using thermocouple arrays. Residual stresses in the as-welded condition and after post-weld heat treatment were measured using diverse methods: neutron diffraction and the contour method at 30 mm thickness; and deep hole drilling and the contour method at 130 mm. Incremental centre hole drilling measurements were performed at 130 mm thickness to better understand near-surface stresses. Weld and heat-affected zone microstructures and microconstituents were evaluated using a combination of hardness mapping, optical microscopy and electron microscopy. The as-welded residual stresses exhibit the characteristic M-shaped distribution for hardenable steels, reaching 500–600 MPa in tension in both steels at both thicknesses. However, the modest changes to the chemical composition and the change in plate thickness both significantly influenced microstructures, mechanical properties and residual stress distributions. These sensitivities underline the need for physically faithful models. This extensive characterisation study enables the development and validation of models that predict the development of microstructures and residual stresses in EB welds in low alloy pressure vessel steels. [ABSTRACT FROM AUTHOR]