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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.

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]

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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]

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.

In situ irradiation in transmission electron microscope following pre-irradiation was performed on Fe-0.6Ni alloy to investigate the growth of dislocation loops under irradiation. The observations showed two types of loop growth behaviour. One type was almost linear growth with irradiation dose, possibly caused by loop absorbing freely migrating defects. The other was abrupt growth with rapid increase in diameter. It was found that nanometre-sized loop might migrate one-dimensionally, following which the large loop could absorb the small loop. This process can result in the abrupt growth of loop and broadening of loop size distribution. Image, graphical abstract [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.

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]

• 9 chemical contents and 3 irradiation conditions were used as input factors. • The probability density distribution of Δ RT NDT was used to evaluated the model performance. • The GBDT model has the best performance, its prediction is close to that of ASTM E900-15. • FNF, T and the content of Cu, Ni, C are the most important 5 factors for irradiation embrittlement. A reliable irradiation embrittlement model of reactor pressure vessel (RPV) is critical to the safe operation and life extension of nuclear power plants (NPP). In this study, 474 datasets from NUREG/CR-6551 and 6 machine learning (ML) methods were adopted to accurately predict the irradiation embrittlement behavior (Δ RT NDT) of RPV with comprehensive input factors involving irradiation conditions and metal composition. The GBDT model has the best prediction performance. The calculated result by GBDT model is close to those of JEAC-4201 and ASTM E900-15. Cu content is the most sensitive factor for Δ RT NDT , followed by neutron fluence. This work demonstrates a successful application of ML in promoting NPP safety and RPV materials design. [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.

Reactor pressure vessel (RPV) of pressurized water reactor undergoes severe neutron irradiation during normal operation and life-extension phase, thus leading to the formation of numerous nanometric irradiation-accelerated and irradiation-induced solute-enriched clusters among the ferritic matrix, including so-called late blooming phases (LBP), which was believed as Mn-Ni-Si enriched stable precipitates to emerge lately and grow rapidly and show great threat to the mechanical performance of RPV steels. However, the mechanism of nucleation and growing for LBP remains not in light, especially the effect of inherent defects such as dislocation and grain boundary on that. We investigated the atomic configurations of Fe-Cu, Fe-Mn, Fe-Ni, Fe-Mn-Ni, Fe-Mn-Cu, Fe-Ni-Cu and Fe-Mn-Ni-Cu as RPV model alloys at 600 K, implemented firstly by constructing an enough big simulation box containing edge dislocation and random-distributed alloy atoms and secondly evolving to equilibrium status after essential simulation steps with Metropolis Monte Carlo (MMC) method. The number of clusters, average solute numbers contained in each cluster and the volume atomic fraction of all clusters were characterized, as well as the structure of typical big clusters among simulation analyzed. [ABSTRACT FROM AUTHOR]

Neutron irradiation following thermal ageing experiments were performed using Fe-Cu, Fe-Cu-Ni, Fe-Cu-Ni-Mn, and Fe-Cu-Si alloys to investigate the effects of Ni, Mn, and Si on the formation of Cu-rich clusters and their chemical composition. Atom probe tomography showed that Ni and Mn increased the number density of Cu-rich clusters. This can be caused by that Ni and Mn reduce the formation energy of Cu nuclei, thus enhancing the Cu nucleation rate. Mn may also enhance the heterogeneous nucleation of Cu because of its strong interaction with self-interstitial atoms. In addition, Mn can accelerate solute diffusion during thermal ageing. A high Ni alloying content enhanced the Mn concentration in the clusters as well as the Ni concentration, which is believed to be caused by the thermodynamic relationship between Ni and Mn. In Fe-Cu-Si alloys, the Si concentration in the clusters was very low and was comparable to the Si alloying content. This indicates that Si atoms hardly cluster with Cu. [ABSTRACT FROM AUTHOR]

A theoretical study is conducted to evaluate the nucleation free energy of copper-vacancy clusters in Fe as a function of the numbers of copper atoms and of vacancies in a cluster. Using this free energy value, cluster nucleation processes during irradiation are investigated. The results clearly show that there are two different types of cluster nucleation paths on the free energy surface; one is the formation of empty voids by jumping over the ridge of the free energy surface, and the other corresponds to a path for the formation of copper clusters by going around the ridge. The dependence of easy nucleation paths on the damage rate is discussed. [ABSTRACT FROM AUTHOR]

A multiscale modeling approach was applied to investigate the nucleation process of CRPs (copper rich precipitates, i.e., copper-vacancy clusters) in α-Fe containing 1 at.% Cu during irradiation. Monte-Carlo simulations were performed to investigate the nucleation process, with the rate theory equation analysis to evaluate the concentration of displacement defects, along with the molecular dynamics technique to know CRP thermal stabilities in advance. Our MC simulations showed that there is long incubation period at first, followed by a rapid growth of CRPs. The incubation period depends on irradiation conditions such as the damage rate and temperature. CRP’s composition during nucleation varies with time. The copper content of CRPs shows relatively rich at first, and then becomes poorer as the precipitate size increases. A widely-accepted model of CRP nucleation process is finally proposed. [ABSTRACT FROM AUTHOR]

Solute cluster induced hardening plays a dominant role in the irradiation embrittlement of reactor pressure vessel. For Cu-rich clusters, the Russell and Brown model (1972) is often applied for their correlation with hardening, but many parameters of the model are taken as adjustable to fit specific experimental data. In this work, a correlation is attempted and discussed between Cu-rich clusters and hardening in a series of reactor pressure vessel model alloys. Compared with the Cu clusters of the same size, Cu-rich clusters containing Ni and Mn may have lower obstacle strength and cause lower hardening. [ABSTRACT FROM AUTHOR]