WWER Reactor Pressure Vessels Research Articles

Порівняння майстер кривої з нормативним методом оцінки в'язкості руйнування металу корпусу реактора ВВЕР-1000

The paper considers the generalized temperature dependence of the stress intensity factor KJC in the WWER-1000 reactor pressure vessel metal, which was derived based on the surveillance test data for fracture toughness. The consistency of the shape of the Master curve and 95 % upper and lower confidence bounds with the experimental dependence of crack resistance parameters on temperature was studied. From the point of view of the level of conservatism and acceptability, a comparison of the normative curves of fracture toughness according to NPRR (Nuclear Power Rules and Regulations) G-7-002-86 and SOU NNEGC (Standard of organization of Ukraine, National Nuclear Energy Generating Company) 177:2019 with a 5 % confidence bound of the Master curve was carried out. The possibility of using RT0 temperature as a temperature index instead of TK for the normative fracture toughness curve of SOU NNEGC 177:2019 was considered.

Improved determination of fast neutron fluence onto the WWER pressure vessel metal surveillance specimens

The special methodology for the determination of the neutron fluences onto the surveillance specimens of the WWER-1000 reactor pressure vessel metal has developed by INR NASU specialists and is successfully applied. The methodology is based on the Monte-Carlo code that is used for neutron transport calculations to the locations of the surveillance specimens. The methodology improvement is described. The fundamentals of the technique of the calculation-experimental determination of the fast neutron fluences onto surveillance specimens and their errors are presented. It is shown a dosimetry experiment in a particular reactor is needed to obtain experimental data about irradiation conditions in the locations of the surveillance specimens.

Estimation of yield strength due to neutron irradiation in a pressure vessel of WWER-1000 reactor based on the correction of the secondary displacement model

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K–P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.

Improvement of Fracture Resistance Calculation Procedure for NPP Primary Equipment Components

The procedure of fracture resistance calculation for WWER primary equipment components has been improved. In particular, this refers to the reactor pressure vessel (RPV) and steam generators (SG) under normal operating conditions and emergencies. The developed calculation procedures and software make it possible to determine the significant effect of such factors as deformation stress history, residual technological heredity, analysis of temperature dependence of stress intensity factors for the postulated crack, regularity and density of the finite element mesh in the crack front on the assessment of WWER-1000 RPV fracture resistance.
 The paper proposes the methodology for justifying the place and orientation of the postulated crack to obtain the most conservative assessment of fracture resistance in the area of RPV inlet nozzles. It is shown that elastoplastic calculations in the thermal shock simulation can help to improve estimates of RPV strength and lifetime. 
 It was established that not taking into account the elastoplastic deformation history, residual technological stresses after heat treatment and corrosion effects can result in non-conservative assessment of fracture resistance of coolant header welding to SG PGV-1000M shell under normal operating conditions and emergencies.
 The calculation methodology and software for assessing stress-strain state of in-vessel internals were improved taking into account state-of-the-art approaches to modeling of radiation-induced swelling deformations and dependence of metal mechanical peculiarities on exposure doses and temperature.

Annealing as a Technique for Estimating the Structural Elements Contribution to NPP Materials Service Properties

The annealing technique is used to estimate the effect of radiation-induced structural elements on the service properties of the materials used in nuclear power plants: reactor pressure vessel (RPV) and in-vessel internals steels. The WWER-1000 RPV weld seam is affected by a hardening mechanism, the contribution of which is 60–65%, and a nonhardening mechanism resulting in grain-boundary segregation, the contribution of which is 30–35%. Radiation defects 40–45% and G-phase precipitates 35–40% are mainly responsible for the radiation-induced hardening of WWER-1000 internals materials. Radiation-induced structural changes should be eliminated for these materials to be reused.

Impact of shield elements on the WWER-440 reactor pressure vessel activation

The WWER-440 type nuclear reactor is one of the most common reactor types in the nuclear power plants (NPP) across Central and Eastern Europe. Most of the NPPs equipped with these reactors have reached or are approaching the end of their service lifetime and face decommissioning. Neutron activated components of these reactors (the largest being the reactor pressure vessel (RPV)) are the most hazardous radioactive waste arising during decommissioning. Some of these reactors were equipped with shield elements (dummy fuel assemblies) in order to reduce neutron flux affecting the RPV. Therefore, the induced activity of an RPV also depends on whether or not shield elements were used in a specific reactor. This paper presents the analysis of shield elements’ influence on the induced activity of an RPV.

A Setup for Complex Investigation of Mechanical Characteristics of Structural Materials for NPP Equipment

The complex experimental equipment has been developed and manufactured in compliance with the existing standards for measuring hardness and strength characteristics via the method of instrumented indentation, as well as for short-term static uniaxial tensile tests of structural materials and small punch testing of miniature disk specimens in the macrorange of loads of 2–10,000 N. This allows one to obtain the necessary characteristics of hardness and strength in relation to the current state of materials of the main equipment at NPP and thereby minimize the volume of material used for the production of specimens on the same setup employing various normative techniques of both destructive and non-destructive inspections. The equipment is designed in the form of replaceable units: indentation, tension, and punching of miniature disk specimens, which are installed depending on the test method. As a result, there is a unified assembly that makes it possible to reduce the effect of instrument errors in the validity of the data obtained using different test methods and facilitate its modification according to the other tests. The developed experimental equipment has been tested using the segment of metal of the double ball pipe wall ∅990 × 70 mm (the basic metal is 10GN2MFA steel and the deposited material is 08Kh19N10G2B steel) of the Du850 main circulation pipeline of NPP from WWER-1000 reactor pressure vessel. The mechanical characteristics obtained by the methods of indirect control were compared with those obtained from standard tensile tests. At the same time, the deviation of the values of yield strength and ultimate strength obtained employing the indirect control methods from the values obtained in the standard tensile testing is within the limits recommended by the normative documents.

Effect of neutron irradiation hardening of the base metal on the results of WWER-1000 reactor pressure vessel residual lifetime assessment

This paper discusses the assessment of the WWER-1000 reactor pressure vessel integrity and the analyses performed to protect the irradiated vessel against brittle failure. The paper focuses on studies assessing the effects of base metal hardening on fast fracture evaluation results. It is shown that the results depend significantly on the material model used (elastic or elastic–plastic) as well as on the base metal mechanical properties (including or not the hardening produced by irradiation). Considering the change in tensile properties of the base material under irradiation may change the identification of the limiting combination of the pressurised thermal shock scenario, the limiting crack front point, and the crack aspect ratio. To demonstrate these effects, the results of the calculations performed for the South Ukrainian Nuclear Power Plant Unit 2 reactor pressure vessel residual lifetime are presented.

A Data Scatter for a Shift of the Ductile to Brittle Transition Temperature for WWER-1000 Reactor Pressure Vessel Materials

In the most countries that operate the nuclear power plants with reactor pressure vessels a safety margin accounting a data scatter is applied for a conservative evaluation of a radiation shift of the ductile to brittle transition temperature for RPV metal. This scatter is to a significant extent due to material inhomogeneity and errors in determining the temperature shift and neutron fluence. In the regulatory practice of Ukraine, the obsolete approaches are used that can lead to an underestimation or overestimation of the transition temperature shift depending on the number of test data points.
 In order to use the updated regulatory approaches that will be consistent with international practice, it is necessary to know the magnitude of the data scatter on the transition temperature shift which is characterized by a standard deviation. Therefore, the aim of the research work was to estimate the data scatter for WWER reactor pressure vessel materials using statistical methods.
 The paper presents the results of a statistical analysis for a large array of surveillance test data for WWER-1000 reactor pressure vessels of NPP units which are operated in Ukraine. The data scatter for RPV base and weld metal has been estimated using a statistical treatment for the dependencies of a transition temperature shift, ΔTF, on the fast (Е > 0,5 MeV) neutron fluence. The ΔTF values have been derived from the Charpy impact tests. The Charpy V-notch specimens have been irradiated in the nuclear power reactors within a neutron fluence range of (3,0 ÷ 92,2)·1022 m-2 in the frame of a national surveillance program.
 The analysis has shown the data scatter relative to the average regression line for RPV materials is characterized by a standard deviation of 5,5 °С. Based on the results obtained, it was suggested to use a double standard deviation of 11 °С as a safety margin to provide a conservative estimate for the radiation shift of the transition temperature of the WWER-1000 reactor pressure vessel materials.

Radiation embrittlement of support structure materials for WWER RPVs

Radiation embrittlement of materials of support structures (SS) for reactor pressure vessel (RPV) of WWER types is considered. The features of radiation embrittlement of SS materials are connected with low irradiation temperature no exceeding 90 °C and also with a use of the steels which are usually applied for different building structures and have not a high resistance to the radiation embrittlement.The experimental date on standard mechanical properties and fracture toughness have been obtained for various steels and weld metals in the initial and irradiation conditions. SEM investigation of fracture surface of fractures toughness specimens and investigation with atomic-tomography method have been performed. The model for prediction of material radiation embrittlement as a function of the neutron fluence and the impurity (Cu and P) content is developed. The irradiation temperature effect is considered on various mechanisms of radiation embrittlement of a material.

Methodology of Including Spatial Nonuniformity of Material Properties in the Brittle Fracture Resistance Calculation for WWER-1000 Reactor Pressure Vessels. Part 2. Experimental Investigations

Using experimental data the authors determine standard deviations of distribution of the ductile-tobrittle transition temperature TK for the base metal (in the vessel shell) and weld metal for WWER-1000 reactor pressure vessels. Based on the standard deviations of TK , defect concentration and size distribution, dimensions of a postulated defect, and on the data regarding the reactor pressure vessel loading during the emergency cooldown, the authors find the values of the margin for spatial nonuniformity in terms of TK , T0, and Ω.

Methodology of Including Spatial Nonuniformity of Material Properties in the Brittle Fracture Resistance Calculation for WWER-1000 Reactor Pressure Vessels. Part 1. Theoretical Analysis

The paper addresses the main methods of predicting the temperature dependence of fracture toughness. The authors have analyzed a system of margins which have to be introduced in order to adequately and conservatively predict this dependence using the results of testing surveillance coupons. A methodology is put forward for determining a margin for spatial nonuniformity of properties of the reactor pressure vessel material, which is based on simultaneous deterministic and probabilistic calculations of brittle fracture resistance of a reactor pressure vessel.

Iron Filtered Neutron Spectra: Measurement and Comparison with Calculations Using Different Libraries

Iron filtered neutron spectra were measured for benchmark assemblies of three types: Fe spheres of 30, 50 and 100 cm diameter with a 252Cf neutron source placed in the center; WWER-1000 reactor pressure vessel mock-up (Fe slab 20 cm); and Fe slabs 67 and 120 cm thick in the reactor beam. Monte Carlo simulations with the MCNP-4c code using different evaluated data libraries ENDF/B-VII.1, BROND-3, JENDL-4.0, JEFF-3.2T2, CENDL-3.1 and TENDL-2012 were performed for spherical assemblies focusing on the neutron energy range of 0.06 to 1.3 MeV. Considerable discrepancies with experimental data were observed for all libraries.

Prediction of the effects of thermal ageing on the embrittlement of reactor pressure vessel steels

A new method has been proposed for prediction of the effects of thermal ageing on the embrittlement of reactor pressure vessel (RPV) steels. The method is based on the test results for materials in two conditions, namely, aged at temperatures of temper embrittlement and annealed after irradiation. The prediction is based on the McLean’s equation and the dependencies describing thermally activated and radiation-enhanced phosphorus diffusion. Experimental studies have been carried out for estimation of thermal ageing of the WWER-1000 RPV 2Cr–Ni–Mo–V steel. The ductile to brittle transition temperature shift ΔTk due to phosphorus segregation has been estimated on the basis of experimental data processed by the proposed method for the time t=5×105h (more than 60years of operation) for the base and weld metals of the WWER-1000 RPV.

Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs during Operation (VERLIFE)

VERLIFE – “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation “was developed within the 5th Framework Programme of the European Union in 2003 and later upgraded within the 6th Framework Programme “COVERS – Safety of WWER NPPs” of the European Union in 2008.This Procedure has to fill the gap in original Soviet/Russian Codes and Rules for WWER type NPPs, as these codes were developed only for design and manufacture and were not changed since their second edition in 1989.VERLIFE Procedure is based on these Russian codes but incorporates also new developments in research, mainly in fracture mechanics, and also some principal approaches used in PWR codes.To assure that VERLIFE Procedure will remain a living document, new 3-years IAEA project (in close co-operation with another project 6th Framework Programme of the European Union “NULIFE – Plant Life Management of NPPs”) started in 2009.Within this project, upgrading/updating of the VERLIFE procedure is prepared together with the extension by (at least) following procedures:-Lifetime of reactor pressure vessel internals-Leak-before-break concept for WWER NPPs-No-break-zone for WWER NPPs-Probabilistic approach for failure and lifetime assessment of WWER reactor pressure vessels and primary piping (including Risk-informed ISI)-Qualification of NDT procedures-Thermal fatigue evaluation-Supports of components and pipingFinal document, after its approval by expert groups of the IAEA and NULIFE, will be issued in 2014 as “IAEA VERLIFE Guidelines for Integrity and Lifetime Assessment of Components and Piping in WWER NPPs”.

Microstructural study of WWER reactor pressure vessel steels

This paper presents results from our long-term studies of irradiated, commercially used WWER reactor pressure vessel steels. Results from Charpy-V tests and positron annihilation spectroscopy techniques are compared and discussed in details, having in mind actual state of art and other microstructural studies in this area. The optimal region for annealing of irradiation induced defects was analysed. It was shown that WWER steel with low impurity contents has good radiation stability and operation these reactor pressure vessels could be extended beyond a design lifetime.

Flux effect analysis in WWER-440 reactor pressure vessel steels

The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013–0.036wt%) and copper (0.08–0.20wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ∼4×1012сm−2s−1 and low ∼6×1011сm−2s−1 fluxes (E>0.5MeV) at a temperature of 270°С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25–50°C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13wt%).Microstructural studies show that copper content decreases in the ferrite matrix and that there is formation of copper-enriched clusters and precipitates under irradiation and subsequent thermal annealing. Due to this fact the radiation damage does not depend on copper re-irradiation after annealing. Dependence of transition temperature shift on neutron flux is not expected for WWER-440 steels under re-irradiation after annealing.

Analysis of Relationship Between the Radiation Embrittlement Mechanisms and the Influence of Neutron Flux in Respect of WWER Reactor Pressure Vessel Materials

The paper addresses the effect of neutron flux for WWER reactor pressure vessel materials in the cases where various embrittlement mechanisms prevail. The data on shifts of critical brittleness temperatures, which were obtained within various surveillance specimen programs and research programs, are used. The authors have defined the embrittlement mechanisms for which the neutron flux has no effect and the ones for which its effect is significant. For the case where the phosphorus segregation mechanism dominates, the absence of the neutron flux effect has been substantiated from the standpoint of the theory of radiation-accelerated diffusion.

Elastoplastic Fracture Resistance Analysis of NPP Primary Circuit Equipment Elements

Nurrent calculation analysis approaches to the fracture resistance assessment for main equipment elements of the WWER NPP primary circuit are considered. It is noted that the solution of the problems of justifying safe operation, assessment of the integrity and extension of the lifetime of the reactor pressure vessel is essentially dependent on the results of calculations of the kinetics of the stress strain state, a proper consideration of the residual welding stress and strain fields, reliable evaluation of non-linear fracture mechanics parameters. A general methodology is developed for a calculated analysis of the fracture resistance of WWER reactor pressure vessels in the simulation of emergency cooling conditions. The main statements in the elastoplastic calculation of the kinetics of the stress strain state for reactor pressure vessels are formulated with consideration of the residual welding stress and strain fields, and the calculation procedure for determining the fracture parameters of the postulated crack is presented. Using the calculation methods and software product developed, a significant influence of such factors as consideration of the thermomechanical loading history and elastoplastic deformation of the metal in the vicinity of the postulated crack front, different ways of taking into consideration the residual welding stress level, the regularity and density of the finite element mesh in the vicinity of the calculated crack front, the procedure for calculating the elastoplastic fracture parameters in discrete finite element method models on the calculated fracture resistance assessment of WWER-1000 reactor pressure vessels is shown. It is found that the neglect of the loading history and residual stress level after heat treatment results in a non-conservative assessment of the fracture resistance for the collector-to-nozzle welded joint in the PGV-1000 steam generator in the simulation of the operating cycle of loading.

Interpretation of Accelerated Irradiation Results for Materials of WWER-1000 Reactor Pressure Vessels

Studies of WWER-1000 reactor vessel weld materials with different Ni and Mn content after thermal exposure at operating temperature for up to 10,000 h have been performed. It is shown the hold time interval under study is not sufficient for development of thermal ageing processes.