A new approach to description of in-service embrittlement of WWER-1000 reactor pressure vessel materials

Based on available published data and experimental findings, the in-service embrittlement of materials has been analyzed. The contributions of thermal ageing and neutron irradiation to embrittlement of the base and weld metals of WWER-1000 reactor pressure vessels are discussed. Equations have been derived which describe the shift of the critical brittle temperature depending on the irradiation time and neutron fluence. For a nickel-rich weld metal a relation between the radiation embrittlement coefficient and the amount of alloying elements Ni, Mn, and Si has been defined.     

Radiation Embrittlement Understanding for PLIM Activities at EC-JRC-IE

Radiation embrittlement and aging mechanisms for NPP reactor pressure vessels and vessel internals have been studied within NPP Plant Life Management activities for the evaluation, prediction, and monitoring of the critical components' service life. The main achievements of the SAFELIFE project, integrating various networks on PLIM issues, are given. Results of neutron embrittlement of model alloys are presented, and surveillance and research data on WWER reactor pressure vessel and other steels are analyzed. Projects for the development of destructive and non-destructive testing of irradiated materials are outlined.     

Probabilistic analysis of mechanical properties of WWER-1000 reactor pressure vessel materials

The variance of strength and plastic characteristics of 15Kh2NMFA steel and its welds is assessed by the results of statistical analysis of material mechanical properties for nineteen WWER-1000 reactor pressure vessels manufactured by Izhorskii and Atommash factories. The values determined with a 95% probability are well within the limits specified by the applicable regulations. The parameters which are needed for the statistical brittle-strength analysis of a reactor pressure vessel are presented with the same veracity. __________ Translated from Problemy Prochnosti, No. 2, pp. 15–28, March–April, 2006.     

Assessment of In-Service Degradation of Mechanical Properties of NPP Reactor and Piping Materials

An experimental investigation has been carried out to clarify the influence of operating factors on static and cyclic strength of NPP reactor vessels and piping. The authors present the results of analysis of crack growth resistance characteristics of the base material and welded joints in WWER-1000 reactors (steel 15Kh2NMFA) in fatigue failure. The mechanisms of thermal aging of steels of NPP piping in long-term operation (up to 10,000 hours) have been studied.     

Beyond RPV Design Life

A set of standard surveillance programs for monitoring the design life of reactor pressure vessels (up to 40 years of operation) has been analyzed. In view of the improved test methods and embrittlement evaluation procedures, the necessity has been shown of introducing modifications in the present surveillance programs aiming at a more precise reactor pressure vessel integrity evaluation facing a possible service life of 60 years. Service life predictions are performed for reactor pressure vessels, based on the available surveillance data, reconstituted Charpy specimens, and Master Curve testing.     

The use of master curve method for statistical re-evaluation of surveillance test data for WWER-1000 reactor pressure vessels

The surveillance fracture toughness test data for WWER-1000 reactor pressure vessel materials from Ukrainian nuclear power plants were re-evaluated using the Master curve methodology. It has been shown that experimental temperature dependence of fracture toughness parameters and a scatter of K_Jc values are in a good agreement with a Master curve shape and 5 and 95% tolerance bounds for materials in unirradiated condition and after neutron irradiation up to fluence 41. 2·10²² n/m² (E > 0.5 MeV). For the Khmelnitsky nuclear power plant unit 1 reactor pressure vessel an analysis has shown that normative approach PNAé G-7-002-86 underestimates essentially the measured fracture toughness of unirradiated weld metal. The reference temperature T₀ calculated according to the Master curve method was compared with a critical brittleness temperature T_K0 for reactor pressure vessel materials in unirradiated condition. It has been found that temperature T₀ is much lower than T_K0 . Furthermore a difference between T₀ and T_K0 values varies essentially from one material to another. A correlation between temperatures T_{28 J} defined from Charpy energy curve and T₀ values calculated from precracked Charpy specimens test was obtained. The analysis has shown that the results based on precracked Charpy specimens can provide nonconservative assessment of fracture toughness for WWER-1000 reactor pressure vessel materials.     

THE MECHANISMS AND DETECTION OF EMBRITTLEMENT IN Cr-Mo PRESSURE VESSEL STEELS

Abstract— 21/4CrlMo steel and 11/4Cr1/2Mo steel have been widely been used for hydro-processing units such as hydro-desulphurising and hydro-cracking reactors. These reactor pressure vessel steels have a potential for temper embrittlement that leads to toughness degradation and a reduction of the critical flaw size for brittle fracture. These steels are also susceptible to hydrogen embrittlement, especially in aged steels where cracks may propagate in the base metal up to the critical flaw size. A vessel with adequate toughness when originally constructed may therefore embrittle during service and such changes may require pressure restrictions during start-up and shut-down.
A survey of the literature shows composition to be the controlling parameter for both temper embrittlement (TE) and hydrogen embrittlement (HE), in-particular the presence of residual impurity elements such as P and the presence of elements such as Mo which nullify the effect of impurity segregation.
Much information is available to describe embrittlement phenomena for Cr-Mo steels. This paper reviews the mechanisms of TE and HE and describes a microstructural characterisation route which subsequently allows the structural integrity of potentially embrittled vessels to be examined for the purposes of remaining life assessment and plant life extension.     

An integrated approach to evaluating the fracture toughness of irradiated nuclear reactor pressure vessels

Current methods of predicting the inservice fracture toughness of nuclear reactor pressure vessels subject to irradiation embrittlement are briefly reviewed, and a new and integrated approach is proposed. This approach is based on the use of a wide variety of information, including the rapidly emerging understanding of the fundamental mechanisms of fracture in the ductile to brittle transition region as well as the microstructurally-mediated processes leading to embrittlement. However, the focus is on advanced, nonintrusive characterization methods for measuring composition, coarse and fine scale microstructure, and mechanical properties using small sample biopsies from operating vessels.     

Radiation effects in nuclear reactor materials—correlation with structure

A review of radiation effects in nuclear reactor materials has been made; the irradiation effects have been correlated with the crystal structure of the materials. Five phenomena, irradiation hardening, irradiation embrittlement, irradiation creep, irradiation growth and void swelling that occur in materials by neutron irradiation in a reactor environment have been discussed with a view to explaining the physics of the phenomena and the engineering consequences. Metallurgical approaches for improving the irradiation performance of materials and for developing new alloys with better resistance to radiation damage have been pointed out.     

Measurement and calculation of the fast-neutron and photon spectra from the core boundary to the biological shielding in the WWER-1000 reactor model

The fast-neutron and photon space-energy distributions have been measured in an axially (1.25 m active height) and azimuthally (60 degree symmetry sector) shortened model of the WWER-1000 reactor assembled in the LR-0 experimental reactor. The space-energy distributions have been calculated with the stochastic code MCNP and the deterministic three-dimensional code TORT. Selected results are presented and discussed in the paper. This work has been done in the frame of the EU 5th FW project REDOS REDOS, Reactor Dosimetry: Accurate determination and benchmarking of radiation field parameters, relevant for reactor pressure vessel monitoring. EURATOM Programme, Call 2000/C 294/04). All geometry and material composition data of the model as well as the available experimental data were carefully checked and revised.     

Multicriteria materials selection for extreme operating conditions based on a multiobjective analysis of irradiation embrittlement and hot cracking prediction models

A methodology for evaluating different combinations of materials specifications for extreme environment applications is presented. This new approach addresses the materials selection problem using a multicriteria stringency level methodology that defines several thresholds obtained by analyzing different prediction models of irradiation embrittlement and hot cracking. To solve the conflicts among thresholds as provided by the different prediction models, a multiobjective approach is carried out. Materials for reactor pressure vessels have been considered as case study. It has been concluded that the best option to manufacture a pressure vessel for a pressurized water modern reactor is the selection of German manufacturing standards. Finally, a sensitivity analysis of the proposed methodology has been performed to evaluate the divergences between the single stringency level methodology and the new proposal including multicriteria decision making aspects.     

Transferability of Laboratory Fracture Data to Safety Assessment of Postulated Defects in Critical NPP Components

For strength evaluation of reactor pressure vessels, large-scale benchmark tests have been conducted on determination of fracture toughness of specimens with internal defects. The obtained results provide a good check on the transferability of data obtained on standard materials to the assessment of real and postulated flaws in reactor pressure vessels. The Master Curve method makes it possible to get more reliable predictions of ductile-brittle transition, in comparison with the approaches using K_Ic as a function of RT_NDT.     

Brittle fracture local criterion and radiation embrittlement of reactor pressure vessel steels

The application of local criteria for predicting brittle fracture of reactor pressure vessel steels is discussed with an emphasis on radiation embrittlement. An association of the radiation-induced damages and the processes of initiation and propagation of cleavage microcracks is analyzed from the standpoint of the local criterion for fracture. Physical-mechanical models are put forward to describe the influence of radiation damages on the cleavage microcrack initiation. The influence of the material hardening caused by neutron irradiation and plastic deformation on the fracture toughness is studied.     

Charpy impact test on A508-3 steel after neutron irradiation

Reactor pressure vessel (RPV) is the critical un-changeable component of the reactor during its service lifetime, which prevents the radioactive leak of the nuclear power plant core. The irradiation test (about 10 × 10¹⁹ cm^− 2, E > 1 MeV) in research reactor of the pressure vessel material was carried out, and the charpy impact test has been carried out before and after the neutron irradiation. Analysis of the impact energy and the fracture morphology has been done to estimate the embrittlement due to neutron irradiation. It shows that the main effects of neutron irradiation to fracture are the crack initiation and stable expansion process. And there also are a small amount of intergranular fracture in the unstable crack expansion after neutron irradiated which aware us pay more attention to the increasing intergranular fracture behavior of higher neutron fluence level after 60a nuclear power plant operation.     

Evolution of the structure of the WWER-1000 reactor vessel from the prototype model to WWER-1200

We discuss the guidelines used in the design of the vessels of WWER-1000 (water-cooled water-moderated) nuclear reactors. The characteristic specific features of the structure of the reactor vessel are discussed, and the subsequent steps of improvement of its design aimed at the increase in its operating safety and designed service life up to 60 years are proposed.     

Radiation-Induced Embrittlement of WWÉR-440 Reactor Pressure Vessel Steel under Loading

The results of evaluating the reference temperature T ₀ are presented and Master Curve is constructed from the experimental data for steel 15Cr2MoVA (base metal of WWÉR-440 reactor pressure vessel) in three states: unirradiated, irradiated, and irradiated under stress. It is shown that mechanical load, which simulates the coolant pressure, accelerates radiation embrittlement, whose contribution is comparable to that of neutron irradiation.     

Finite Element Analysis of Hydrogen Transport in Steel Pressure Vessel at Room Temperature

Hydrogen embrittlement is commonly considered as an important failure mechanism for steel pressure vessels and pipes made of such as Cr–Mo and 4130X steels under high-pressure hydrogen environments, which means hydrogen atom can easily penetrate and diffuse into the metal, leading to the distortion of microscopic lattice and the degradation of macroscopic strength and fracture toughness. Under the support of the National Key Fundamental Research and Development Project of China (2015.1-2019.12), we aim to launch a series of theoretical, experimental, and numerical research on the macroscopic damage evolution and microscopic fracture of steel structures under high-pressure hydrogen environment, which ultimately commits to gaining deep insight into the hydrogen embrittlement mechanisms. This work studies the hydrogen transport mechanisms in Cr–Mo steel pressure vessels under different hydrogen environments using finite element analysis (FEA), which is fundamental to subsequent research on the hydrogen-induced damage evolution and crack behaviors. The purpose of this paper is to explore the effects of the initial hydrogen concentrations and structural sizes on the hydrogen transport mechanisms in 2.25Cr-1Mo pressure vessels with a nozzle at room temperature. Numerical results by comparing different hydrogen concentration distributions show that structural discontinuities tend to accelerate the hydrogen embrittlement sensitivity.     

The influence of thermomechanical loading history on the stress level in WWER NPP reactor pressure vessels under thermal shock

We present the results of calculations of the kinetics of stress-strain state and stress intensity factors for surface and under-the-cladding circumferential cracks in modeling the emergency core cooldown conditions for the WWER-1000 reactor. The calculation procedure is based on a mixed finite-element method statement which provides stability of numerical solution and a high accuracy of results for both the displacements as well as stresses and strains. The authors analyze the influence of the density of the finite-element discretization of the crack-tip area for the surface and under-the-cladding circumferential cracks on the accuracy and convergence of computation of fracture-mechanics parameters in the modeling of thermal shock conditions. The results of calculation of kinetics of stress intensity factors allowing for the thermomechanical loading history and residual process-induced stress fields are reported. It is demonstrated that if the elastoplastic deformation history and residual process-induced stress fields are disregarded in the calculations of stress intensity factors for under-the-cladding cracks the reactor pressure vessel strength and lifetime may turn out to be overestimated.     

Study of the beltline weld and base metal of WWER-440 first generation reactor pressure vessel

We present experimental results of the circumferential core weld SN0.1.4 and the base metal ring 0.3.1 of the reactor pressure vessel from the Unit 1 of the Greifswald WWER-440/230. The investigated trepans represent the irradiated–annealed– reirradiated (IAI) condition. The working program is focused on the characterization of the reactor pressure vessel steels through the reactor pressure vessel wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921 to determine the fracture toughness in different thickness locations.