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The reactor pressure vessel (RPV) in onshore nuclear power plants is typically analysed for fatigue life by considering the temperature, internal pressure, and seismic effects using a simplified time-domain fatigue analysis. In contrast, the frequency-domain fatigue analysis method is commonly employed to assess the fatigue life of ship structures. The RPV of a floating nuclear power plant (FNPP) is subjected to a combination of temperature, internal pressure, and wave loads in the marine environment. Consequently, it is essential to effectively integrate the frequency-domain fatigue analysis method used for hull structures with the time-domain fatigue analysis method for RPVs in FNPPs or, alternatively, to develop a suitable method that effectively accounts for the temperature, internal pressure, and wave loads. In this study, a quasi-time-domain method is proposed for the fatigue analysis of RPVs in FNPPs. In this method, secondary components of marine environmental loads are filtered out using principal component analysis. Subsequently, the stress spectrum induced by waves is transformed into a stress time history. Fatigue stress under the combined influence of temperature, internal pressure, and wave loads is then obtained through a stress component superposition method. Finally, the accuracy of the quasi-time-domain method was validated through three numerical examples. The results indicate that the calculated values obtained by the quasi-time-domain method are slightly higher than those obtained by the traditional time-domain method, with a maximum deviation of no more than 24%. Additionally, the computation time of the quasi-time-domain method is reduced by 98.67% compared to the traditional time-domain method. [ABSTRACT FROM AUTHOR]

A computational route was developed for precise calculation of fast neutron fluence on a WWER-type reactor pressure vessel (RPV). The method is based on the transfer of neutronics data from HELIOS-2 lattice calculations and nodal diffusion neutronics data (power, density, and temperature) from BIPR7.1 and PARCS 3.36/PATHS core calculations into a three-dimensional (pinwise axially distributed) fixed neutron source for modeling of transport of fast neutrons from the reactor core to the outer surface of the RPV using MCNP6.2. Validation of the proposed computational method was carried out based on comparative analysis of MCNP6.2-predicted and neutron dosimetry–measured reaction rates [54Fe(n,p)54Мn, 93Nb(n,nʹ)93mNb, and 58Ni(n,р)58Со] on the outer surface of the Armenian Nuclear Power Plant (ANPP) Unit 2 RPV. Validation revealed that the MCNP6.2-predicted fast neutron fluence results are very sensitive to the ENDF-B neutron data. Particularly, MCNP6.2 with ENDF/B-VIII.0 significantly underpredicts (20% to 30%) fast neutron fluence while using ENDF/B-VII.1 data overpredicts it. Adding revised beta-released evaluations of 54Fe, 56Fe, 57Fe, and 16O from the International Nuclear Data Evaluation Network (INDEN) to ENDF/B-VIII.0 allows one to obtain reasonable agreement with measurement results for all types of measured reaction rates. [ABSTRACT FROM AUTHOR]

The precipitation of solutes is a major cause of irradiation hardening and embrittlement limiting the service life of reactor pressure vessel (RPV) steels. Impurities play a significant role in the formation of precipitation in RPV materials. In this study, the effects of carbon on cluster formation and irradiation hardening were investigated in an RPV alloy Fe-1.35Mn-0.75Ni using C and Fe ions irradiation at 290 °C. Nanoindentation results showed that C ion irradiation led to less hardening below 1.0 dpa, with hardening continuing to increase gradually at higher doses, while it was saturated under Fe ion irradiation. Atom probe tomography revealed a broad size distribution of Ni–Mn clusters under Fe ion irradiation, contrasting a narrower size distribution of small Ni–Mn clusters under C ion irradiation. Further analysis indicated the influence of carbon on the cluster formation, with solute-precipitated defects dominating under C ion irradiation but interstitial clusters dominating under Fe ion irradiation. Simulations suggested that carbon significantly affected solute nucleation, with defect clusters displaying smaller size and higher density as carbon concentration increased. The higher hardening at doses above 1.0 dpa was attributed to a substantial increase in the number density of defect clusters when carbon was present in the matrix.

This study addresses cracks and fracture problems in engineering structures that may cause significant challenges and safety concerns, with a focus on pressure vessels in nuclear power plants. Comprehensive parametric three-dimensional mixed mode fracture analyses for inclined and deflected nozzle corner cracks with various crack shape aspect ratios and depth ratios in nuclear reactor pressure vessels are carried out. Stress intensity factor (SIF) solutions are obtained using FRAC3D, which is part of Fracture and Crack Propagation Analysis System (FCPAS), employing enriched finite elements along the crack front. Also, improved empirical equations are developed to allow the determination of mixed mode SIFs, KI, KII, and KIII, for any values of the parameters considered in the study. This study provides practical solutions to assess the remaining life and fail-safe conditions of nuclear reactors by providing accurate SIF determination.

The microstructure and hardness of fusion boundary (FB) and near FB of a low alloy steel (LAS) 16MND5 welding cladded with stainless steel 308L for nuclear reactor pressure vessel were investigated by using modern electron microscopes and nano-indenter. The wavy morphology of fusion line results in changes of nearby microstructures of both cladding 308L and substrate 16MND5 along the FB. The microstructure of the cladding 308L is composed of columnar grains adjacent most of the FB, but it is equiaxed grains at the valley of the wavy fusion line with large curvature. The microstructures of 16MND5 near the FB are ferrite at the wavy fusion line with large curvature, martensite at the wavy fusion line with small curvature, and transition mixed structure between them. A carbide-depleted zone adjacent to the FB was observed in the 16MND5 side, of which the hardness is lower than both FB and 16MND5 base metal. Three types of FB can be observed according to the width, being narrow or sharp FB, needle-like martensite FB, and partially-mixed zone FB. The FB includes high density of carbides of M7C3 and M3C, leading to a hardness peak at FB.

The reactor pressure vessel (RPV) in onshore nuclear power plants is typically analysed for fatigue life by considering the temperature, internal pressure, and seismic effects using a simplified time-domain fatigue analysis. In contrast, the frequency-domain fatigue analysis method is commonly employed to assess the fatigue life of ship structures. The RPV of a floating nuclear power plant (FNPP) is subjected to a combination of temperature, internal pressure, and wave loads in the marine environment. Consequently, it is essential to effectively integrate the frequency-domain fatigue analysis method used for hull structures with the time-domain fatigue analysis method for RPVs in FNPPs or, alternatively, to develop a suitable method that effectively accounts for the temperature, internal pressure, and wave loads. In this study, a quasi-time-domain method is proposed for the fatigue analysis of RPVs in FNPPs. In this method, secondary components of marine environmental loads are filtered out using principal component analysis. Subsequently, the stress spectrum induced by waves is transformed into a stress time history. Fatigue stress under the combined influence of temperature, internal pressure, and wave loads is then obtained through a stress component superposition method. Finally, the accuracy of the quasi-time-domain method was validated through three numerical examples. The results indicate that the calculated values obtained by the quasi-time-domain method are slightly higher than those obtained by the traditional time-domain method, with a maximum deviation of no more than 24%. Additionally, the computation time of the quasi-time-domain method is reduced by 98.67% compared to the traditional time-domain method.

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. [ABSTRACT FROM AUTHOR]

This study uses computational fluid dynamics (CFD) to investigate the threedimensional flow field under normal operating conditions in the reactor pressure vessel (RPV) in the Hualong One nuclear power plants (NPPs). With a particular focus on the flowrate distribution at the core inlet, the numerical framework is validated against the integral hydraulic experiment in a 1:4-scaled RPV of CNP1000, the prototype of the Hualong One reactor. The simulation results of the normalized flowrate at the core inlet agree reasonably well with the measured data. Based on the experimental data, several methods of calibrating the CFD turbulence model coefficients are suggested by introducing the concepts of data assimilation and machine learning. The flow field in a realistic RPV for Hualong One is predicted using the validated numerical framework, showing that the flowrate distribution at the core inlet is nearly homogeneous and that the turbulent intensity is acceptably low for each fuel assembly. It can provide essential information for the reactor core thermal–hydraulic design and the fuel assembly mechanical assessment. [ABSTRACT FROM AUTHOR]

It is necessary to quantify the effects of flux on reactor pressure vessel steel embrittlement under neutron irradiation, if surveillance or high-flux test reactor data is used to predict vessel embrittlement occurring at lower fluxes. This is particularly important when considering embrittlement occurring during extended (60-80 years) operation for which there is no direct experience. Dedicated investigations are time-consuming and expensive even when only small flux-fluence ranges are investigated, so collating data from multiple campaigns is necessary to provide sufficient information to cover the wide range of fluxes required for vessel assessment in the long term. This paper collates and reviews such data. The review finds that flux dependences probably differ in sign and strength in different regimes (low flux and fluence, intermediate flux at low and high fluence, high flux at low and high fluence) with the regime limits affected by composition and temperature. The current understanding of diffusion processes and microstructural development are invaluable in interpreting the trends and limits. Many contradictory data sets were found, however, and not all contradictions could be dismissed as resulting from poor quality data. Suggestions are made for investigations to clarify the uncertainties. One wide-ranging model of flux effects, based on an extensive data set, is used to compare high-fluence data from different sources, to assess whether embrittlement rates accelerate after a high, threshold fluence. The model helps to identify experiments which investigated comparable fluxfluence-temperature regimes. The comparable data are split evenly between data sets supporting acceleration after a particular fluence and data sets contradicting it. The model identifies regimes in which further campaigns would clarify the causes of these contrasting observations. [ABSTRACT FROM AUTHOR]

Copyright of Computer Measurement & Control is the property of Magazine Agency of Computer Measurement & Control and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

In reactor pressure vessel materials, the formation of Mn- and Ni-rich nanoclusters is a major cause of neutron irradiation embrittlement. The segregation of these solute atoms into dislocation loops has attracted attention as a mechanism to accelerate solute clustering. In this study, the behaviors of solute Mn and Ni atoms in Fe-0.6wt.%Ni, Fe-1.4wt.%Mn, and Fe-1.4wt.%Mn-0.6wt.%Ni alloys irradiated at 400 °C up to 3 dpa were analyzed using three-dimensional atom probe tomography. Solute atom clusters were observed in all materials, and their shapes were spherical, flat, and torus in FeNi, FeMn, and FeMnNi, respectively. In ternary alloy FeMnNi, Mn and Ni atoms were concentrated in the sample in the form of arcs, and the orientation of the plane containing the arcs was estimated by comparing field desorption images. The size, number density, and orientation of this structure were found to be in good agreement with those of both types of dislocation loops, namely, b = 1/2 〈111〉 and b = 〈100〉, identified in a previous study using the same material. The positions of Ni and Mn enrichment did not fully overlap. Ni atoms tended to be concentrated more in the inner part of the loop than the Mn atoms. Mn atoms were enriched only in the vicinity of the dislocation loops, whereas Ni atoms showed a higher concentration inside the dislocation loops than in the bulk.

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.

This study uses computational fluid dynamics (CFD) to investigate the three-dimensional flow field under normal operating conditions in the reactor pressure vessel (RPV) in the Hualong One nuclear power plants (NPPs). With a particular focus on the flowrate distribution at the core inlet, the numerical framework is validated against the integral hydraulic experiment in a 1:4-scaled RPV of CNP1000, the prototype of the Hualong One reactor. The simulation results of the normalized flowrate at the core inlet agree reasonably well with the measured data. Based on the experimental data, several methods of calibrating the CFD turbulence model coefficients are suggested by introducing the concepts of data assimilation and machine learning. The flow field in a realistic RPV for Hualong One is predicted using the validated numerical framework, showing that the flowrate distribution at the core inlet is nearly homogeneous and that the turbulent intensity is acceptably low for each fuel assembly. It can provide essential information for the reactor core thermal–hydraulic design and the fuel assembly mechanical assessment.

It is necessary to quantify the effects of flux on reactor pressure vessel steel embrittlement under neutron irradiation, if surveillance or high-flux test reactor data is used to predict vessel embrittlement occurring at lower fluxes. This is particularly important when considering embrittlement occurring during extended (60–80 years) operation for which there is no direct experience. Dedicated investigations are time-consuming and expensive even when only small flux-fluence ranges are investigated, so collating data from multiple campaigns is necessary to provide sufficient information to cover the wide range of fluxes required for vessel assessment in the long term. This paper collates and reviews such data. The review finds that flux dependences probably differ in sign and strength in different regimes (low flux and fluence, intermediate flux at low and high fluence, high flux at low and high fluence) with the regime limits affected by composition and temperature. The current understanding of diffusion processes and microstructural development are invaluable in interpreting the trends and limits. Many contradictory data sets were found, however, and not all contradictions could be dismissed as resulting from poor quality data. Suggestions are made for investigations to clarify the uncertainties. One wide-ranging model of flux effects, based on an extensive data set, is used to compare high-fluence data from different sources, to assess whether embrittlement rates accelerate after a high, threshold fluence. The model helps to identify experiments which investigated comparable flux-fluence-temperature regimes. The comparable data are split evenly between data sets supporting acceleration after a particular fluence and data sets contradicting it. The model identifies regimes in which further campaigns would clarify the causes of these contrasting observations.

The size distribution of former austenite grains in the metal of the heat-affected zone (HAZ) of a welded joint made of 15Kh2NMFAA steel is studied in detail. Computational simulation of the thermal cycle of welding was carried out in order to evaluate the distribution of the maximum values of the temperature and heating rate in the heat-affected zone. The work is aimed at analyzing the causes of grinding of the austenite grain of the shell metal in the HAZ due to the thermal cycle of welding. [ABSTRACT FROM AUTHOR]

An essential aspect of extending safe operation of the world’s active nuclear reactors is understanding and predicting the embrittlement that occurs in the steels that make up the Reactor pressure vessel (RPV). In this work we integrate state of the art machine learning methods using ensembles of neural networks with unprecedented data collection and integration to develop a new model for RPV steel embrittlement. The new model has multiple improvements over previous machine learning and hand-tuned efforts, including greater accuracy (e.g., at high-fluence relevant for extending the life of present reactors), wider domain of applicability (e.g., including a wide-range of compositions), uncertainty quantification, and online accessibility for easy use by the community. These improvements provide a model with significant new capabilities, including the ability to easily and accurately explore compositions, flux, and fluence effects on RPV steel embrittlement for the first time. Furthermore, our detailed comparisons show our approach improves on the leading American Society for Testing and Materials (ASTM) E900-15 standard model for RPV embrittlement on every metric we assessed, demonstrating the efficacy of machine learning approaches for this type of highly demanding materials property prediction.

For nuclear power generation as a carbon‐neutral energy source, in‐vessel retention (IVR) must be implemented to maintain the structural integrity of nuclear reactor pressure vessel (RPV) for more than 72 h under severe accidental conditions. This technology requires accurate prediction of creep deformation and life of RPV material being operated under pressure and extremely high‐temperature gradient. The current work develops a simplified deformation‐mechanism‐based true‐stress (DMTS) model for creep behavior/life‐prediction of SA508 Gr.3 steel, a typical RPV material, above the phase transformation temperatures (800–1000°C). This model is used to evaluate the time to specific creep strain (t3% and t5%) and rupture (tr), in comparison with popular empirical methods such as Orr–Sherby–Dorn (OSD) and Larson–Miller (LM). The simplified DMTS model achieves an excellent agreement with the experimental observations. The controlling deformation mechanisms are also discussed by metallurgical examinations, which provide the physical premise for the model development and application. Highlights: A simplified deformation‐mechanism‐based true‐stress (DMTS) creep model was proposed.This model predicted the creep‐strain time and life of SA508 Gr. 3 at 800–1000°C.The prediction accuracy of simplified DMTS was compared with LM and OSD methods.The dominant creep deformation mechanism was dislocation climb‐plus‐glide. [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. [ABSTRACT FROM AUTHOR]

In this paper, weld metal from unique material of a decommissioned boiling water reactor pressure vessel is investigated. The reactor was in operation for 23 effective full power years. The elemental distribution of Ni, Mn, Si and Cu in the material is analysed using atom probe tomography. There are no well-defined clusters of these elements in the weld metal. However, some clustering tendencies of Ni was found, and these are interpreted as a high number density of small features. Cu atoms were found to statistically be closer to Ni atoms than in a fully random solid solution. The impact of the non-random elemental distribution on mechanical properties is judged to be limited.

A probabilistic fracture mechanics (PFM) analysis code, PASCAL, has been developed by Japan Atomic Energy Agency for failure probability and failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. To strengthen the applicability of PASCAL, considerable efforts on verifications of the PASCAL code have been made in the past years. As a part of the verification activities, a working group consisted of different organizations from industry, universities and institutes, was established in Japan. In the early phase, the working group focused on verifying the PFM analysis functions for RPVs in pressurized water reactors (PWRs) subjected to pressurized thermal shock (PTS) events. Recently, the PASCAL code has been improved in order to run PFM analyses for both RPVs in PWRs and boiling water reactors (BWRs) subjected to a broad range of transients. Simultaneously, the working group initiated a verification plan for the improved PASCAL through independent PFM analyses by different organizations. Concretely, verification analyses for a PWR-type RPV subjected to PTS transients and a BWR-type RPV subjected to a low-temperature over pressure (LTOP) transient were performed using PASCAL. This paper summarizes those verification activities, including the verification plan, analysis conditions and results. Through the verification studies, it is confirmed that the mathematical models and probabilistic calculation algorithms incorporated in PASCAL work appropriately and the applicability of PASCAL has been improved.

Post-irradiation annealing of neutron-irradiated reactor pressure vessel steels is a matter of both technical and scientific interest. Small-angle neutron scattering (SANS), while being sensitive to nm-sized irradiation-induced solute-atom clusters, provides macroscopically representative and statistically reliable measures of cluster volume fraction, number density and size. In the present study, SANS was applied to uncover the size distribution of clusters in as-irradiated samples of a VVER-1000 weld and their gradual dissolution as function of the post-irradiation annealing temperature. The same samples were used to measure Vickers hardness. The results are consistent with Mn-Ni-Si-rich clusters of less than 2 nm radius to be the dominant source of both scattering and hardening. Annealing gave rise to small but significant partial recovery at 350°C and almost complete recovery at 475°C. The dispersed-barrier hardening model was applied to bridge the gap between the characteristics of nano-features and macro-hardness.

In this paper, the evaluations of metal ablation processes under high temperature, i.e., the Al plate ablated by a laser and a heat carrier and the reactor pressure vessel ablated by a core melt, are studied by a novel peridynamic method. Above all, the peridynamic formulation for the heat conduction problem is obtained by Taylor's expansion technique. Then, a simple and efficient moving boundary model in the peridynamic framework is proposed to handle the variable geometries, in which the ablated states of material points are described by an additional scalar field. Next, due to the automatic non-interpenetration properties of peridynamic method, a contact algorithm is established to determine the contact relationship between the ablated system and the additional heat carrier. In addition, the corresponding computational procedure is listed in detail. Finally, several numerical examples are carried out and the results verify the validity and accuracy of the present method. [ABSTRACT FROM AUTHOR]

To ensure the safety margin of a reactor pressure vessel (RPV) under normal operating conditions, it is regulated through the pressure-temperature (P-T) limit curve. The stress intensity factor (SIF) obtained by the internal pressure and thermal load should be obtained through crack analysis of the nozzle corner crack in advance to generate the P-T limit curve for the nozzle. In the ASME code Section XI, Appendix G, the SIF via the internal pressure for the nozzle corner crack is expressed as a function of the cooling or heating rate, and the wall thickness, however, the SIF via the thermal load is presented as a polynomial format based on the stress linearization analysis results. Inevitably, the SIF can only be obtained through finite element (FE) analysis. In this paper, simple prediction equations of the SIF via the thermal load under, cool-down and heat-up conditions are presented. For the Korean standard nuclear power plant, three geometric variables were set and 72 cases of RPV models were made, and then the heat transfer analysis and thermal stress analysis were performed sequentially. Based on the FE results, simple engineering solutions predicting the value of thermal SIF under cool-down and heat-up conditions are suggested.

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

The safety of a nuclear power plant during incidents, such as a Loss of Coolant Accident (LOCA), heavily relies on conducting a comprehensive structural integrity assessment of both the Reactor Pressure Vessel (RPV) and its components, specifically to withstand Pressurized Thermal Shock (PTS). PTS is characterized by a combination of steep temperature gradient, resulting from the injected emergency coolant during a LOCA, and internal pressure within the RPV. Majority of the reported fracture assessment studies on RPV steel, whether experimental or numerical investigations, have predominantly focused on standard uniaxial specimens at iso-thermal loading conditions. To better simulate the thermal shock scenario in RPVs, the present work aims to assess the impact of a biaxial stress field on both fracture parameters (crack mouth opening displacement and J -integral) and plastic collapse load. This assessment is conducted through experimental and numerical investigations, both with and without prior transient thermal load. Fracture experiments are performed on cruciform bend specimens, featuring two different biaxiality ratios (1:1 and 2:1). Moreover, numerical studies on cruciform specimens are conducted using finite element analysis to validate and corroborate the observations derived from the fracture experiments. • Impact of a biaxial stress on fracture parameters and plastic collapse load is studied under thermal shock scenario. • Fracture experiments conducted on cruciform bend specimens of two different biaxiality ratios under thermal transient • Finite element analysis carried out to validate experimental observations. • Results showed that thermal shock reduces load-carrying capacity. • Equi-biaxial cruciform specimens exhibit a higher level of constraint. [ABSTRACT FROM AUTHOR]

• The turbulent flow was obtained with three kinds of flow mixing devices in lower plenum. • Turbulent coherent structures were extracted to study the turbulent mixing mechanism of flow mixing devices. • High-precision matched-index-of-refraction technique is employed to measure the turbulent flow in a scaled-down reactor pressure vessel. Turbulent flow mixing in lower plenum of reactor pressure vessel (RPV) of pressurized water reactor (PWR) is important for design of flow mixing devices. In this study, turbulent flow in a scaled-down RPV model with three designs of mixing devices, including vortex suppression plate, flow skirt and semi-spherical perforated drum, was measured by time-resolved particle image velocimetry (TR-PIV) with aid of the high-precision MIR technique of sodium iodide (NaI) solution and acrylic. Turbulent coherent structures were studied based on the experimental data. The time-averaged flow structures in lower plenum include flow separation at lower support plate, fluid bifurcation at the flow mixing devices, horizontal jets at holes of flow skirt, flow turning horizontally below the flow mixing devices and flow turning upward through the flow mixing devices. The strong turbulent mixing induced by flow mixing devices improves Reynolds stresses in lower plenum and introduces large spatial–temporal length scales in lower plenum. [ABSTRACT FROM AUTHOR]

• Our experiments demonstrate that flame-sprayed porous coatings can significantly enhance the CHF at different inclination angles. • The in-vessel retention engineering verification test on a full-height RPV bottom head is carried out. • The flame-sprayed porous coating has the potential to improve the safety margin of the reactor pressure vessel. Considering the urgent need for enhancing critical heat flux (CHF) on the outer surface of the reactor pressure vessel (RPV) by in-vessel retention (IVR) on Generation III pressurized water reactors (PWRs), we prepare a porous coating of 316 L stainless steel on the surface of SA508 Grade 3 (SA508–3) steel by flame spraying in this study. Using a small-scale pool boiling experimental setup, we first investigate boiling heat transfer characteristics on bare surfaces and flame-sprayed porous coating surfaces of SA508–3 steel in deionized water at atmospheric pressure, and obtain pool boiling curves and CHFs at different inclination angles. Furthermore, we carry out the IVR engineering verification tests for a full-height RPV bottom head on the REPEC-II test facility built at Shanghai Jiao Tong University. Our experimental results demonstrate that the flame-sprayed porous coatings can significantly enhance the CHF by 21–50 % at different inclination angles of the RPV bottom head. Our full-height facility tests demonstrate that the CHF in the high-inclination angle region, which normally has very high heat flux due to the focusing effect, exceeds 2 MW/m2 on the flame-sprayed porous coating surface. Therefore, our flame-sprayed porous coating has the potential to improve the safety margin of RPV. [ABSTRACT FROM AUTHOR]

In-vessel retention（IVR） is widely used as the mitigation strategy for severe accident in the design of China third-generation nuclear power plants. The key to the success lies of this strategy is the heat dissipation from the outer wall of the reactor pressure vessel（RPV） higher than the decay heat. Under the action of the high temperature corium pool in the reactor, significant thermal expansion of the RPV will occurs and the shape of the cooling channel outside the RPV will be extruded, which may affect the local heat transfer. In order to provide the input conditions for the heat transfer study of cooling channel, a thermal-hydraulic and solid mechanics coupling calculation model was built through the COMSOL multiphysics coupling calculation software to study the thermal expansion and deformation of the RPV under severe accidents. The results show that under the cooling condition of the RPV, the outer cooling channel will be extruded obviously, which should be considered in the engineering design. The internal pressure has an obvious influence on the expansion and deformation of the RPV, depressurizing of the primary circuit after a serious accident is one of the keys for the success of the IVR strategy. [ABSTRACT FROM AUTHOR]

Copyright of Nuclear Safety is the property of Nuclear & Radiation Safety Center and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Radiation damage calculation is very important for estimating the lifetime of nuclear power plant instruments. The purpose of the present study is to evaluate the sensitivity of the damage values to the neutron group energy, where the neutron flux rate is maximum. In the present study, we propose three neutron group energy (WIMS, CINDER, and OPENMC) to evaluate the sensitivity of radiation damage calculations. The obtained result from SPECTER code shows that the differences between these results and the standard values (ASTM E-693) for three group energy WIMS, CINDER and OPENMC are respectively 0.25E-04, 0.75E-04, and 1.74E-04 in the ¼ diameter of reactor pressure vessel (RPV). Therefore, the WIMS energy group is the most accurate spectrum compared to other spectrums in this region. Also, these differences are respectively 0.43E-04, 0.52E-04, and 1.86E-04 in the ¾ diameter of RPV. Moreover, the WIMS group energy is the most accurate spectrum in this region. This result shows that increasing the number of neutron energy group cause to reduce the difference between calculation of damage and its standard values. According to the results of calculated damage that induced by WIMS group energy in ¼ diameter of RPV, the values of yield strength at 1, 5, 10, 15, 20, 25, 30, 35 and 40 years are calculated by SPECOMP and SPECTER codes.

As far as the reactor pressure vessel steel is concerned, the isothermal fatigue study of welded joints is helpful to protect the pressure vessel from blasting and save the life of its operators. Furthermore, the study helps to improve the future design of the reactor components and also enthuse to develop new materials. In material science, the fatigue of metal is quite difficult to deal with, and that of welded joint is an even more complex phenomenon. This complexity arises from different main factors: thermal cycles during welding, which will strongly affect the base material and induce residual stresses and distortion, the fusion process with the filler metal leads to heterogeneity to microstructure which will affect or change the original mechanical properties and chemical composition through out of the welded joint. In this study, isothermal mechanical fatigue of welded reactor steel has been contemplated. The samples of the welded joint used in this study are the '15H2MFA steel type of VVER-440' reactor pressure vessel operating at Paks Nuclear Power Plant, Hungary. The reactor vessel is made of low-alloy heat-resistant steel and austenitic steel with a corrosion-resistant inner (clad) surface. [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.

• Examined added mass effects in concentric cylinders with a new set of narrow gaps. • Revisited theoretical added mass effects in concentric cylinders. • Accurately captured modal parameters with modal analysis and modal testing. • Confirmed added mass effects in narrow gap by tests on cylinders and reactor vessel. Nuclear reactor pressure vessels (RPVs) must be robust to external environmental excitations such as seismic loads. Precise numerical models that predict the dynamic responses to external loads are crucial for evaluating the safety of RPVs. Notably, in RPVs submerged in reactor coolant, it is necessary to examine the added mass effect of the fluid via fluid-structure interaction (FSI) analysis, employing either an analytical approach or a numerical analysis. The present study focuses on modeling and analyzing the added mass effect in the narrow gaps associated with the dense structural layout in the reactor. In this context, this paper validates analytical and numerical models that accurately describe the added mass effect for two cases: (i) simple concentric cylinders with a narrow gap filled with fluid and (ii) a complex small-scale RPV with narrow gaps also filled with fluid. Initially, the analytical model, employing Lagrange's equation of motion and velocity potential, and the numerical model, utilizing FE modeling in acoustic FSI through ANSYS, were validated through experimental modal testing for the added mass effect in the narrow gap between simple concentric cylinders. The analytical model for the concentric cylinders indicates that the added mass diverges to infinity as the gap approaches zero, which should be examined carefully. Through experiments for the concentric cylinders, the analytical and FEM-based added mass effects were validated with a natural frequency error of 3 %. Subsequently, based on the results for the concentric cylinders, a complex numerical model of the small-scale reactor with complex boundary conditions was developed and cross-validated through corresponding experimental modal testing of the small-scale RPV. The added mass model for the narrow gaps in the RPV was validated by substantial agreement of eight representative modes, as demonstrated by a 5 % natural frequency error. [ABSTRACT FROM AUTHOR]

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]

• A novel multiscale simulation strategy linked hydrogen diffusion to the cracking. • Micro-structure dependent non-uniform lattice hydrogen diffusion was examined. • Internal shearing and necking modes induced different failure mechanisms was determined. • The risk of hydrogen-induced failure at different stages of void evolution was evaluated. Leveraging on a recently developed multiscale simulation method, we propose a pioneering strategy for the multiscale analysis of hydrogen-induced cracking in reactor pressure vessels that considers micro-voids, their influences on hydrogen diffusion as well as the operating pressure of the vessel. In the finite element analysis, the nodal lattice hydrogen concentration and displacements are exchanged bi-directionally between the macroscale and microscale within a monolithic solution procedure. During the analysis, the voids are shown to cause non-uniform lattice hydrogen diffusion and the trapping effect, resulting in inter-void hydrogen accumulation, is demonstrated. Based on the hydrogen concentration on the void surfaces, the void hydrogen pressure was calculated for the hydrogen-induced cracking analysis. It was found that both internal shearing and necking failure modes exhibited micro-structure dependence. The internal shearing mode caused oriented through-wall propagation of cracks in the vessel pressure while the internal necking mode promoted void nucleation in the reactor pressure vessel. As a result, during the early stages of void evolution, the oriented internal shearing of the dispersed small voids showed a high possibility to induce through-wall failure of the reactor pressure vessel. [Display omitted] [ABSTRACT FROM AUTHOR]

• Pre-tempering at 200 °C and directly tempering at 650 °C result in a similar microstructure. • Tensile strength of Mn-Ni-Mo steel is not dependent on pre-tempering temperature. • Highest UTS*TE index and UTS values are obtained at different pre-tempering temperatures. • Pre-tempering at 350 °C formed aggregated carbide precipitates of various sizes. • Pre-tempering at 500 °C led to the most favourable microstructure for practical usage. Mn-Ni-Mo steels used in reactor pressure vessel manufacturing often exhibit martensite-austenite (M/A) islands post-quenching. These islands demonstrate the deleterious impact of M/A island decomposition on the mechanical properties of Mn-Ni-Mo steels, as they can decompose into long, rod-shaped Fe 3 C carbides when tempered above 600 °C. To address this issue, the current study investigates a two-step tempering process to gain deeper insight into the characteristics of M/A island decomposition and its impact on the mechanical properties and fracture surface morphology of tensile and impact-tested samples. The process involves a two-hour pre-tempering treatment at 200 °C, 350 °C, and 500 °C, followed by three hours of final tempering at 650 °C. The study reveals temperature-dependent characteristics of M/A island decomposition, as well as the significant impact of tempered microstructure and carbide precipitates (rod-shaped M 3 C, rhombus-shaped M 7 C 3 , and globular-shaped M 23 C 6) on mechanical properties and fracture surface morphology. At 200 °C, the M/A island partially decomposes, resulting in long rod-shaped carbide precipitates during final tempering. These precipitates reduce the critical cleavage stress required for microcrack initiation, resulting in lower ultimate tensile strength and impact toughness. Furthermore, the fracture surface of the sample pre-tempered at 200 °C exhibits a higher percentage of cleavage facets than samples pre-tempered at 350 °C and 500 °C. Meanwhile, at 350 °C, the pre-tempering sample showed aggregated carbide precipitates of various sizes at former M/A island locations, resulting in the formation of numerous micro-voids in close proximity during the fracture. These voids eventually coalesce, forming a substantial crack that leads to material failure. The 500 °C pre-tempering sample contained coarse spherical carbides, and carbide precipitate clustering had disappeared. Samples pre-tempered at 300 °C and 500 °C had a higher fraction of ductile dimples, indicating better tensile properties. However, splitting-related decohesion and dimple depth increased as the pre-tempering temperature increased. All tempered samples show dimples and quasi-cleavage facets in the fracture propagation region, indicating a mixed-mode fracture mechanism. [ABSTRACT FROM AUTHOR]

In a postulated severe accident, the thermo-mechanical loads from the corium debris that has relocated to the lower head of the reactor pressure vessel (RPV) can pose a credible threat to the RPV’s structural integrity. In case of a vessel breach, it is vital to predict the mode and timing of the vessel failure. This affects the ex-vessel accident progression and plays a critical role in the development of mitigation strategies. We propose a methodology to assess RPV failure based on MELCOR and ANSYS Mechanical APDL simulations. A Nordic-type boiling water reactor (BWR) is considered with two severe accident scenarios: i) SBO (Station Blackout) and ii) SBO + LOCA (Loss of Coolant Accident). In addition, the approach considers the dynamic ablation of the vessel wall due to a high-temperature debris bed with the use of the element kill function in ANSYS. The results indicate that the stress failure mechanism is the major cause of the RPV failure, compared to the strain failure mechanism. Moreover, the axial normal stress and circumferential normal stress make the dominant contributions to the equivalent stress σ at the lower head of RPVs. As expected, the region with high ablation is most likely the failure location in both SBO and SBO + LOCA. In addition, comparisons of the failure mode and timing between SBO and SBO + LOCA are described in detail. A short discussion on RPV failure between ANSYS and MELCOR is also presented.

The paper presents the results of the application of the refined calculation guidelines to the analysis of the stress-strain state and fracture resistance of the equipment elements of the NPP primary circuit system with WWER reactor, in particular, the cylindrical part and the zone of the reactor pressure vessel inlet nozzle, coolant collector welded joint in a steam generator, baffle, and reactor internals' shaft under the operating and emergency modes of thermomechanical loading. The calculation methodology is based on the use of the modern models and methods of solution to the nonlinear boundary tasks of non-isothermal thermal viscoplasticity and tasks of fracture mechanics using the refined mixed schemes of the finite element method with increased accuracy. They allow one to obtain reliable solutions to the tasks considering the deformation history of thermomechanical loading, modern concepts of fracture mechanics and approaches to the modeling of brittle-viscous fracture defects, irradiation swelling and creep of metal, as well as as the structural elements, under the conditions of elevated temperature, intensive neutron irradiation and nonstationary modes of thermomechanical loading. The authors obtained and analyzed the calculation results for the cylindrical part and zone of the inlet nozzle of the WWER-1000 reactor pressure vessel, coolant collector welded joint in a steam generator, baffle, and reactor internals' shaft using the proposed guidelines. The results imply the possibilities of the proposed recommendations and make it possible to analyze some general tendencies and features of the application of the refined calculation models to justify the strength and life extension of the NPP structural elements under investigation. [ABSTRACT FROM AUTHOR]

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