The toughness module of the PERFECT platform: A predictive tool for the fracture toughness of RPV steels

The PERFECT project of the 6th Framework Program aims at developing a predictive tool for irradiation effects on Reactor Pressure Vessel steels. In this work, we focuse on the mechanical part of the numerical platform, the Toughness Module. Its main objective is to predict the probability of failure of the considered RPV steel, using more or less complex approaches. Six submodules are integrated in the Toughness Module. Three of them allow to estimate the macroscopic stress-strain curve of the material and the three others allow to predict the toughness drop of the material due to irradiation.     

Effects of hydrazine addition and N2 atmosphere on the corrosion of reactor vessel steels in diluted seawater under gamma-rays irradiation

Seawater was injected into the reactor cores in the Fukushima Daiichi Nuclear Power Station. Corrosion of primary containment vessel (PCV) steel and reactor pressure vessel (RPV) steel is considered to progress until the molten fuel debris is removed. To evaluate durability of the PCV and RPV steels, corrosion tests were conducted in diluted seawater at 50 °C under gamma-rays irradiation of dose rates of 4.4 and 0.2 kGy/h. To evaluate the effect of hydrazine (N₂H₄) as an oxygen scavenger under gamma-rays irradiation, 10 and 100 mg/L N₂H₄ were added to the diluted seawater. Without addition of N₂H₄, weight loss in the PCV and RPV steels irradiated with the 0.2 kGy/h dose rate was comparable with those without irradiation and weight loss in the vessel steels irradiated with the 4.4 kGy/h dose rate was higher than those without irradiation. Under irradiation, weight loss in the PCV and RPV steels in diluted seawater containing N₂H₄ was comparable with that in diluted seawater without N₂H₄. When gas phase in the flask was replaced with N₂, weight loss in the PCV and RPV steels, and O₂ and H₂O₂ concentrations in the diluted seawater decreased.     

RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called ‘Experimental Test Reactors’. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, …) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program.     

Nuclear material investigations by advanced analytical techniques

Advanced analytical techniques have been used to characterize nuclear materials at the Paul Scherrer Institute during the last decade. The analysed materials ranged from reactor pressure vessel (RPV) steels, Zircaloy claddings to fuel samples. The processes studied included copper cluster build up in RPV steels, corrosion, mechanical and irradiation damage behaviour of PWR and BWR cladding materials as well as fuel defect development. The used advanced techniques included muon spin resonance spectroscopy for zirconium alloy defect characterization while fuel element materials were analysed by techniques derived from neutron and X-ray scattering and absorption spectroscopy.     

Component vibration of VVER-reactors — diagnostics and modelling

Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have lead to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostics, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies.     

低铜合金反应堆压力容器钢辐照脆化预测评估模型

反应堆压力容器（RPV）材料辐照脆化预测评估对保证核反应堆安全运行、预防重大灾难性事故的发生具有重要意义。通过深入了解RPV材料辐照损伤机理和分析国外较为成熟的RPV辐照脆化预测模型,揭示了国外有关压力容器辐照脆化预测模型对低铜RPV辐照脆化预测的不足及其原因。在此基础上,发展和建立了适用于低铜RPV辐照脆化趋势的预测模型CIAE-2009。利用辐照性能数据对CIAE-2009模型进行了验证。结果表明,CIAE-2009对低铜含量RPV材料辐照脆化趋势预测具有较高的准确性和可靠性。     

Comparison of microstructural features of radiation embrittlement of VVER-440 and VVER-1000 reactor pressure vessel steels

Comparative microstructural studies of both surveillance specimens and reactor pressure vessel (RPV) materials of VVER-440 and VVER-1000 light water reactor systems have been carried out, following irradiation to different fast neutron fluences and of the heat treatment for extended periods at the operating temperatures. It is shown that there are several microstructural features in the radiation embrittlement of VVER-1000 steels compared to VVER-440 RPV steels that can cause changes in the contributions of different radiation embrittlement mechanisms for VVER-1000 steel.     

TEM and PAS study of neutron irradiated VVER-type RPV steels

Conventional transmission electron microscopy and positron lifetime and Doppler broadening positron annihilation spectroscopy techniques have been used to investigate the radiation-induced microstructural changes in surveillance specimens of VVER-type reactor pressure vessel (RPV) steels, and RPV steels irradiated in the research reactor. Defects visible in transmission electron microscopy consist of black dots, dislocation loops and precipitates concentrated along the dislocation substructure. Their size and density depend on the neutron flux and fluence. The parallel set of thermally aged specimens, specimens recovery annealed after irradiation and specimens irradiated in a lower neutron flux was investigated too. No defects discernible in transmission electron microscopy were found after accelerated irradiation in the research reactor. In addition to visible defects, the small-volume vacancy clusters were identified by positron annihilation spectroscopy.     

Microstructural aspects of neutron embrittlement of reactor pressure vessel steels - A view from positron annihilation spectroscopy

A comprehensive review of positron annihilation studies of Cr---Mo---V reactor pressure vessel (RPV) steels (Soviet type 15Kh2MFA) in unirradiated and neutron irradiated states is presented. The influences of lattice defects, impurity atom distribution, irradiation temperature, flux and fluence of fast neutrons on positron annihilation parameters, especially during isochronal annealing, are discussed in terms of the positron trapping model. In contrast to the literature, where irradiation-enhanced Cu precipitates and solute coated microvoids are thought to be major defect types responsible for strengthening and hence embrittling of RPV steels, we suggest irradiation-induced precipitates, i.e. probably carbides, to play this role. Possibilities to probe this model are suggested.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Current understanding of radiation-induced degradation in light water reactor structural materials

Current phenomenological knowledge and understanding of mechanisms are reviewed for radiation embrittlement of reactor pressure vessel low alloy steels and irradiation assisted stress corrosion cracking of core internals of stainless steels. Accumulated test data of irradiated materials in light water reactors and microscopic analyses by using state-of-the-art techniques such as a three-dimensional atom probe and electron backscatter diffraction have significantly increased knowledge about microstructural features. Characteristics of solute clusters and deformation microstructures and their contributions to macroscopic material property changes have been clarified to a large extent, which provide keys to understand in the degradation mechanisms. However, there are still fundamental research issues that merit study for long-term operation of reactors that requires reliable quantitative prediction of radiation-induced degradation of component materials in low-dose rate high-dose conditions.     

Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels

Positron annihilation spectroscopy (PAS) and a computer simulation were used to investigate a defect production in reactor pressure vessel (RPV) steels irradiated by neutrons. The RPV steels were irradiated at 250 °C in a high-flux advanced neutron application reactor. The PAS results showed that mainly single vacancies were created to a great extent as a result of a neutron irradiation. Formation of vacancies in the irradiated materials was also confirmed by a coincidence Doppler broadening measurement. For estimating the concentration of the point defects in the RPV steels, we applied computer simulation methods, including molecular dynamics (MD) simulation and point defect kinetics model calculation. MD simulations of displacement cascades in pure Fe were performed with a 4.7 keV primary knock-on atom to obtain the parameters related to displacement cascades. Then, we employed the point defect kinetics model to calculate the concentration of the point defects. By combining the positron trapping rate from the PAS measurement and the calculated vacancy concentrations, the trapping coefficient for the vacancies in the RPV steels was determined, which was about 0.97 × 10¹⁵ s⁻¹. The application of two techniques, PAS and computer simulation, provided complementary information on radiation-induced defect production.     

Towards a modelling of RPV steel brittle fracture using crystal plasticity computations on polycrystalline aggregates

In the frame of the multi-scale approach of the fracture toughness prediction defined in the PERFECT project, we proposed a new crystal plasticity model and applied it to the computation of stress heterogeneities within a reference polycrystalline aggregate defined in the project RPV material.The proposed crystal plasticity model is able to take into account the effects of temperature and irradiation hardening. The analysis of the results of aggregate computations shows that the distributions of the maximum values of the maximal principal stresses are found to be well described by a Gumbell function. Applying these distributions on a Griffith criterion allows settling the basis of an original fracture criterion. However the increasing resistance to fracture of the steel with temperature can be reproduced only by introducing a temperature dependence of the fracture energy.     

Small-angle neutron scattering study on the effect of hydrogen in irradiated reactor pressure vessel steels

Hydrogen uptake can enhance the neutron embrittlement of reactor pressure vessel (RPV) steels. This suggests that irradiation defects act as hydrogen traps. The evidence of hydrogen trapping was investigated using the small-angle neutron scattering (SANS) method on four RPV steels. The samples were examined in the unirradiated and irradiated states and both in the as-received condition and after hydrogen charging. Despite the low bulk content of hydrogen achieved after charging with low current densities, an enrichment of hydrogen in small microstructural defects could be identified. Preferential traps were microstructural defects in the size range of ≈ > 10 nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.     

Evaluation of irradiation damage effect by applying electric properties based techniques

The most important effect of the degradation by radiation is the decrease in the ductility of the pressure vessel of the reactor (RPV) ferritic steels. The main way to determine the mechanical behaviour of the RPV steels is tensile and impact tests, from which the ductile to brittle transition temperature (DBTT) and its increase due to neutron irradiation can be calculated. These tests are destructive and regularly applied to surveillance specimens to assess the integrity of RPV. The possibility of applying validated non-destructive ageing monitoring techniques would however facilitate the surveillance of the materials that form the reactor vessel.The JRC-IE has developed two devices, focused on the measurement of the electrical properties to assess non-destructively the embrittlement state of materials. The first technique, called Seebeck and Thomson Effects on Aged Material (STEAM), is based on the measurement of the Seebeck coefficient, characteristic of the material and related to the microstructural changes induced by irradiation embrittlement. With the same aim the second technique, named Resistivity Effects on Aged Material (REAM), measures instead the resistivity of the material.The purpose of this research is to correlate the results of the impact tests, STEAM and REAM measurements with the change in the mechanical properties due to neutron irradiation. These results will make possible the improvement of such techniques based on the measurement of material electrical properties for their application to the irradiation embrittlement assessment.     

Reactor pressure vessel design of the high temperature engineering test reactor

The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc.2 1/4Cr–1Mo steel is chosen as the material for RPV. The temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 10¹⁷ n/cm² (E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J-factor and are used.In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J-factor and , and nil-ductility transition temperatures, T_NDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described.     

Characterization of neutron-irradiated ferritic model alloys and a RPV steel from combined APT, SANS, TEM and PAS analyses

Understanding the behavior of reactor pressure vessel (RPV) steels under irradiation is a mandatory task that has to be elucidated in order to be able to operate safely a nuclear power plant or to extend its lifetime. To build up predictive tools, a substantial experimental data base is needed at the nanometre scale to extract quantitative information on neutron-irradiated materials and to validate the theoretical models. To reach this experimental goal, ferritic model alloys and French RPV steel were neutron irradiated in a test reactor at an irradiation flux of 9 × 10¹⁷ nm⁻² s, doses from 0.18 to 1.3 × 10²⁴ nm⁻² and 300 °C. The main goal of this paper is to report the characterization of the radiation-induced microstructural change in the materials by using the state-of-the-art of characterization techniques available in Europe at the nanometre scale. Possibilities, limitations and complementarities of the techniques to each other are highlighted.