VVER-1000 Reactor Pressure Vessel Research Articles

Small-angle neutron scattering study of neutron-irradiated and post-irradiation annealed VVER-1000 reactor pressure vessel weld material

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.

Phase Formation Features of Reactor Pressure Vessel Steels with Various Ni and Mn Content under Conditions of Neutron Irradiation at Increased Temperature

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 1022 n/m2 at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (>1.5 wt.%) and Mn (>0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ2-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C).

Precipitation kinetics of radiation-induced Ni-Mn-Si phases in VVER-1000 reactor pressure vessel steels under low and high flux irradiation

• VVER-1000 RPV steels after low and high flux irradiation are studied by APT, TEM. • Number density of NMS phases increases with irradiation dose and Ni+Mn content. • Сomposition of NMS phases does not depend on the irradiation conditions. • High flux irradiation leads to the lower size and volume fraction of NMS phases. One of the key embrittlement mechanisms in reactor pressure vessel (RPV) steels is the hardening produced by nanometer features. In this paper low Cu high Ni VVER-1000 RPV steels with a wide range of Ni (1.1–1.94wt.%) and Mn (0.38–1.1wt.%) contents irradiated at 290°C to (6–101) 10 22 n/m 22 (E>0.5MeV) at both low (surveillance specimens) and high (test reactor) fluxes were analyzed using atom probe tomography and transmission electron microscopy. Formation of high number density of NMS precipitates with the average composition close to at.%: 45Ni-33Si-22Mn regardless of the irradiation conditions was observed. Its number density increases with the increase of the total Ni and Mn concentration in steel. High flux irradiation demonstrates lower size and volume fraction of NMS precipitates compared to the low flux irradiation in case of welds. For the base metal specimens with low Ni (1.1–1.3 wt.%) and Mn (0.38–0.51) content no flux effect was observed. The contribution of NMS phases and dislocation loops into radiation hardening of RPV steels was established by the dispersed barrier strength model (Orowan equation).

Influence of Ni-Mn contents on the embrittlement of PWR RPV model steels irradiated to high fluences relevant for LTO beyond 60 years

Understanding the irradiation induced degradation of the reactor pressure vessel (RPV) steels at high fluences becomes more important in view of safe long term operation (LTO) of existing nuclear power plants. Specifically, the role of Ni, Mn and Si and associated synergetic effects on embrittlement behaviour need to be further investigated to understand accelerated embrittlement at high neutron fluences (ca. 1 × 10 20 n.cm −2 , E>1 MeV) observed in certain types of RPV steels. In addition, it is required to verify the validity of existing embrittlement trend curves (ETCs) at high fluence values. To address these issues, NRG and JRC performed a unique irradiation experiment called LYRA-10 in the High-Flux Reactor (HFR) Petten. In LYRA-10 a variety of specimens (e.g. tensile, Charpy, mini-Charpy and microscopy) of RPV steels (base metal and welds) with systematic variations in Ni, Mn and Si contents, while keeping the base composition representative of VVER-1000 and PWR RPV steels, have been irradiated for 16 HFR cycles to achieve the nominal fast neutron fluence of 1.11 × 10 20 n.cm −2 at a nominal temperature of 286 °C. A large post-irradiation examination campaign on LYRA-10 specimens has been initiated within the STRUMAT-LTO, which is an international collaboration project co-funded by EURATOM. The research reported in this paper includes results from tensile and mini-Charpy impact testing of four PWR RPV model steel specimens irradiated in LYRA-10, which is performed as part of STRUMAT-LTO in-kind pilot project before the start of the EURATOM co-funded project. The combined influence of Ni and Mn on embrittlement behaviour, characterized by irradiation hardening and shifts in transition temperature (T k ) properties, has been evaluated. The measured shifts in T k values for all four model steels have been compared with predicted values from two selected ETCs in order to check their validity for these steels at the above high neutron fluence values.

Investigation of the VVER-1000 reactor pressure vessel neutron fluence and displacement per atom using MCNP6

Ensuring the integrity of the reactor pressure vessel (RPV) during the operating life of nuclear reactors is of paramount importance, and hence the harmful effects of irradiation on the steels of the RPVs have been extensively studied especially the irradiation embrittlement effects. However, improvements in the calculation methods are still necessary to further reduce the discrepancy between calculations and experiments. This study was performed to examine the neutron fluence and respective rate of displacement per atom (DPA), an important parameter describing the irradiation-induced damage, in the VVER-1000 RPV steel based on the criticality calculation, instead of the common fixed source approach, with the MCNP6 Monte Carlo code, and identify the most sensitive locations in the VVER-1000 RPV. The MCNP6 modeling of the VVER-1000 reactor was verified against the OpenMC calculations in addition to the confirmation of the convergence of the effective neutron multiplication factor and the Shannon entropy of the fission source distribution in the MCNP6 criticality calculations to ensure reliable tally results of the neutron fluence. Consequently, the maximum neutron fluence in the VVER-1000 RPV was found at the positions adjacent to the peripheral fuel assemblies, and the respective DPA rate was analyzed in correlation with the neutron fluence and energy spectrum.

Short-term strength properties and features of high-temperature deformation of VVER reactor pressure vessel steel 15Kh2NMFA-A within the temperature range 20–1200 °C

The problem discussed in this work concerns the estimate of integrity of the VVER reactor pressure vessel in the severe accident. In particular, the main goal of discussed investigation is to determine short-term mechanical properties of ferritic-pearlitic vessel steel 15Kh2NMFA-A (2% Cr, 1% Ni, 0.5% Mn, 0.5% Mo) within the temperature range 20–1200 °C. This steel type is one of basic structure steel types used in fabrication of reactor pressure vessels of Russian design, particularly, the VVER vessels. To perform short-term tests of this steel, several lots of steel specimens 8 and 10 mm in diameter with gauge part of length 40–50 mm, correspondingly, were made. The test specimens were made of the fragment of unirradiated standard VVER-1000 reactor pressure vessel of Russian design fabricated by OMZ Group “Izhorskiye zavody” (Russia). The short-term tensile tests of steel specimens were performed both in air within the temperature range 20–1100 °C and in vacuum within the temperature range 900–1200 °C. Basing on the results of the tests, strength-and plastic/strain parameters of this steel (ultimate tensile strength, yield stress, tensile Young’s modulus, steel elongation at fracture, the coefficient of linear thermal expansion) were determined. The analysis of obtained results made it possible to reveal a series of singularities of 15Kh2NMFA-A steel strain. In particular, it was shown that in the temperature range 770–820 °C (Ac1-Ac3), the investigated steel undergoes inner phase/structure transformations, which have an essential influence on strength- and strain properties of investigated steel. The analysis of temperature effect on strength properties (ultimate tensile strength and yield stress) and Young’s modulus of this steel made it possible to reveal three temperature ranges in which these parameters have characteristic features of change. In the first of these ranges (the temperature values from 20 to ~650 °C), a moderate degradation of strength- and elastic properties of steel is observed. In the second range (from ~650 to 850 °C), a sharp decrease of values of steel strength and elasticity occurs, and at test temperature of more than 900 °C, a monotonous decrease of these mechanical properties is observed. The comparison of the results obtained in the present work with the earlier results of tests at temperature of more than 900 °C showed their satisfactory agreement.

Correlation of local parameters of brittle crack initiation and propagation with the fracture toughness of VVER-1000 pressure vessel steels

Base metal of the VVER-1000 reactor pressure vessel steel in the unirradiated state was studied. The purpose of the work was to find a correlation between local fracture microparameters and macroparameters of samples under quasi-static loads. Fracture toughness tests of steel were performed in the transition temperature region. Structural parameters and brittle fracture origin types were defined by means of fractographic analysis of the tested samples. The stress-strain state of the studied samples at brittle fracture initiation moment was calculated using the complex experimental-calculative method.

Contribution of Hardening Mechanism to VVER-1000 RPV Welds Flux Effect

.

Study of Current State of Templates Cut Out of VVER-440 Reactor Pressure Vessel

.

Phase and structural transformations in VVER-440 RPV base metal after long-term operation and recovery annealing

This study was carried out to evaluate the possibility of 1st generation VVER-440 reactors lifetime extension by recovery re-annealing with the respect to base metal (BM). Comprehensive studies of the structure and properties of BM templates (samples cut from the inner surface of the shells in beltline region) of operating VVER-440 reactor (after primary standard recovery annealing 475 °C/150 h and subsequent long-term re-irradiation within reactor pressure vessel (RPV)) were conducted. These templates were also subjected to laboratory re-annealing 475 °C/150 h.TEM, SEM and APT studies of BM after laboratory re-annealing revealed significant recovery of radiation-induced hardening elements (Cu-rich precipitates and dislocation loops). Simultaneously a process of strong phosphorus accumulation at grain boundaries occurs since annealing temperature corresponds to the maximum reversible temper brittleness development. The latter is not observed for VVER-440 weld metal (WM).Comparative assessment of the properties return level for the beltline BM templates after recovery re-annealing 475 °C/150 h showed that it does not reach the one typical for beltline WM after the same annealing.

Evaluation of radiation damage in belt-line region of VVER-1000 nuclear reactor pressure vessel

Regarding irradiation damage, the most sensitive region in the reactor pressure vessel (RPV) is the area adjacent to the reactor core; the so called belt-line region. Belt-lines and their related heat affected zones (HAZs) in the belt-line region are constantly under neutron and gamma irradiation and have a higher potential to various flaws. The microscopic defects produced in materials due to irradiation are referred to as radiation damage; the units of displacements per atom (DPA) are used routinely to characterize the extent of such damage. In this research, for a typical VVER-1000 (V446) RPV, the areas of the highest neutron radiation flux (belt-line) is first identified by using a Monte Carlo neutron transport code. This is used to calculate and validate the neutron flux and fluence at the belt-line. Employing the PTRAC output from MCNP and the SRIM Monte Carlo code, radiation damage in the RPV is simulated and assessed in terms of DPA. The integrated damage for a full effective year is evaluated to be order of 5E-04 displacements per cm thickness of RPV belt-line metal.

Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50–400)°C

This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50–400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔTK) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects – dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔTK shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔTK shift in the studied range of irradiation temperature and fluence.

Calculation of neutron and gamma fluences on VVER reactor pressure vessel

Embrittlement is one of the most important effects affecting reactor pressure vessel (RPV) aging. RPV is irradiated with neutrons and gammas, especially fast neutrons, which mainly lead to embrittlement of RPV during operation lifetime of nuclear reactors. Therefore, the radiation-induced embrittlement of the RPV should be carefully evaluated. In this paper, a preliminary calculation was performed using the MCNP5 code to identify the areas in the RPV of the VVER-1000/V320 reactor where the neutron and gamma fluxes are maximum. Also, the neutron and gamma fluence distributions on the RPV were investigated and evaluated along with their energy spectra. These calculations are the starting point for the evaluation of radiation damage to RPV of VVER reactors.

Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (TK) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in TK shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the TK shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime.

Applicability of LR-0 mock-up results to VVER-1000 reactor pressure vessel issues

Evaluation of neutron fluence in a reactor pressure vessel (RPV) together with surveillance specimen programs for RPV materials are one of the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of RPV residual lifetime. This paper covers transport of neutrons through the RPV of a VVER-1000 nuclear reactor. This problem is of increased importance as it concerns issues around VVER NPP life extension. With regards to the construction (reduced thickness of the lateral reflector), this issue plays greater role in VVER reactors than in Western types of PWR reactors.RPV material degradation depends mainly on neutron flux and spectra. Both quantities can be calculated or measured. This paper compares MCNP calculations and measurements on zero-power experimental reactor LR-0 with TORT calculations for VVER-1000. The goal is to find a reasonable method for precise estimation of neutron fluence and attenuation factor through the RPV. The calculations were performed with MCNP stochastic code and TORT deterministic transport code. The measurements were performed in a VVER-1000 mock-up placed in reactor LR-0 (Research Center Řež).

Scale factor of standard and mini Charpy specimens from VVER-1000 RPV materials

A database has been developed to establish a correlation between standard and mini-Charpy specimens with the scope of tests of 38 sets of specimens. A methodical procedure has been suggested and implemented which permits a twofold reduction of the scatter of experimental data as compared with a similar approach to VVER-440 reactor pressure vessel (RPV) materials.

Assessment of segregation kinetics in water-moderated reactors pressure vessel steels under long-term operation

In reactor pressure vessel (RPV) bcc-lattice steels temper embrittlement is developed under the influence of both operating temperature of ∼300 °C and neutron irradiation. Segregation processes in the grain boundaries (GB) begin to play a special role in the assessment of the safe operation of the RPV in case of its lifetime extension up to 60 years or more. The most reliable information on the RPV material condition can be obtained by investigating the surveillance specimens (SS) that are exposed to operational factors simultaneously with the RPV itself. In this paper the GB composition in the specimens with different thermal exposure time at the RPV operating temperature as well as irradiated by fast neutrons (E ≥ 0.5 MeV) to different fluences (20–71)·1022 m−2 was studied by means of Auger electron spectroscopy (AES) including both impurity and main alloying elements content. The data obtained allowed to trace the trend of the operating temperature and radiation-stimulated diffusion influence on the overall segregants level in GB. The revealed differences in the concentration levels of GB segregants in different steels, are due to the different chemical composition of the steels and also due to different grain boundary segregation levels in initial (unexposed) state. The data were used to estimate the RPV steels working capacity for 60 years. The estimation was carried out using both the well-known Langmuir-McLean model and the one specially developed for RPV steels, which takes into account the structure and phase composition of VVER-1000 RPV steels, as well as the long-term influence of operational factors.

Validation of 3D Code KATRIN For Fast Neutron Fluence Calculation of VVER-1000 Reactor Pressure Vessel by Ex-Vessel Measurements and Surveillance Specimens Results

Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.

Low cycle thermomechanical fatigue of VVER-440 reactor pressure vessel steels: Investigation the fatigue kinetics and development of a life assessment model

The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed, Ti-stabilized austenitic steel) of the VVER-440 reactor pressure vessels was investigated under fully reversed total strain controlled low cycle fatigue tests. The measurements were carried out in isothermal conditions at 260 °C and with thermal-mechanical conditions in temperature range of 150 to 270 °C using the GLEEBLE 3800 servo-hydraulic thermal-mechanical simulator.Owing the nominal fatigue lifetime for different testing conditions interrupted fatigue tests were carried out to investigate the kinetics of the fatigue evolution. Microstructural evaluation of the samples was performed using transmission electron microscopy as well as X-ray diffraction, and measurement of the dislocations was completed. The course of dislocation density in relation to cumulative usage factor was similar for both materials. However, the nature and distribution of dislocations were different in the individual steels and this resulted in different mechanical behaviours.Using scanning electron microscopy the crack shapes and fracture surfaces were observed and analysed. Crack propagation was assessed in relation to the actual crack size and the loading level.A new low cycle fatigue criterion is presented based on the stored energy, which accumulates in the material during fatigue loading. The new damage parameters are based on the assumption that only the stored part of the introduced energy causes the changes in the microstructure. The proposed model is physically consistent and its prediction accuracy is higher than by the classical strain amplitude and strain energy based approaches.The low cycle fatigue behaviour investigated with the developed engineering model can provide a reference for the remaining life assessment and possible operation life extension of nuclear power plant components.

VVER Knowledge Preservation and Transfer within the Frame of CORONA Project Activities

The CORONA project is funded by the European Commission under the FP7 programme with overall objective to establish a Regional Centre of Competence for VVER Technology and Nuclear Applications. The Centre will provide support and services for preservation and transfer of VVER-related nuclear knowledge as well as know-how and capacity building. Specific training schemes aimed at nuclear professionals and researchers, non-nuclear professionals and students are developed and implemented in cooperation with local, national and international training and educational institutions. Pilot trainings are executed for each specific target group to assess the applicability of the training materials. The training scheme implemented for nuclear professionals and researchers is focussed on the NPP Lifetime Management. The available knowledge on enhancing safety and performance of nuclear installations with VVER technology is used in the preparation of the training materials. The Online Multimedia Training Course on VVER Reactor Pressure Vessel Embrittlement and Integrity Assessment, developed by the joint effort of JRC-IET and IAEA is used in the training. The outcome collected from the trainees showed that the tool meets its primary goal of consolidating the existing knowledge on the VVER RPV Embrittlement and Integrity Assessment, provides adequate ground for transfer of this knowledge.