Irradiated Reactor Pressure Vessel Steels Research Articles

Study of a neutron irradiated reactor pressure vessel steel by X-ray absorption spectroscopy

Reactor pressure vessel (RPV) reference steel samples submitted to neutron irradiations followed by thermal annealing were investigated by X-ray absorption fine structure (XAFS) spectroscopy. Several studies revealed that Cu and Ni impurities can form nanoclusters. In the unirradiated sample and in the only-irradiated sample no significant clustering is detected. In all irradiated and subsequently annealed samples increases of Cu and Ni atom densities are recorded around the absorber. Furthermore, the density of Cu and Ni atoms determined in the first and second shells around the absorber is found to be affected by the irradiation and annealing treatment. The comparison of the XAFS data at Cu and Ni K-edges shows that these elements reside in arrangements similar to bcc Fe. However, the local irradiation damage yields vacancy fractions which were determined from the analysis of XAFS data with a precision of ∼5%.

The Clustering of Cu Atoms in Neutron Irradiated Reactor Pressure Vessel Steels Studied by Positron Annihilation

Defect studies of neutron-irradiated Cr-Mo-V (VVER-440) type reactor pressure vessel steels were performed in the present work. The steels were irradiated in the nuclear power plant reactor under the conditions of a regular operation. Characterization of the irradiation induced defects was performed by two complementary techniques of positron annihilation spectroscopy: (i) positron lifetime spectroscopy was used for identification of defects and determination of defect densities, (ii) coincidence Doppler broadening was employed for investigation of Cu atom aggregates. Long range diffusion of Cu atoms is assisted by the irradiation induced vacancies. The solute Cu atoms form small clusters in the irradiated steels. Subsequent isochronal annealing of the irradiated steel leads to vacancy assisted clustering of Cu atoms and formation of small precipitates. The Cu clusters exhibit maximum diameter at 400oC. Above this temperature the clusters dissolve again in the matrix.

Phosphorus Segregation and Intergranular Embrittlement in Thermally Aged and Neutron Irradiated Reactor Pressure Vessel Steels

Abstract A study on phosphorus (P) segregation at grain boundaries and intergranular embrittlement of reactor pressure vessel steels due to thermal aging and neutron irradiation has been performed for A533B steel plates and weld metals doped with different bulk levels of P. The initial P concentration at grain boundaries of these materials determined by scanning Auger microscopy ranged from 9 to 33 %. The materials thermally aged at temperatures ranging from 450 to 550°C for up to 10000 h exhibited increases in P concentration described by the McLean’s equation. A linear correlation was obtained between P concentration and Charpy 41 J transition temperature (T41J). Fracture toughness tests in the ductile-to-brittle transition temperature region have been conducted by the Master Curve method using 1T-CT specimens. The material subject to intergranular embrittlement exhibited the dual fracture toughness distributions typically observed in inhomogeneous materials. This was probably due to two types of initiation of cleavage and intergranular fracture. The lower tail analysis based on the Master Curve method enabled conservative fracture toughness estimates by fitting the lower fracture toughness arising from intergranular fracture. The relation between median KJc values estimated from the lower tail analysis and temperature appeared to follow the Master Curve. For neutron irradiated materials up to a fluence of 6.9×1019 n/cm2 (E>1 MeV) at 290°C by the Japan Materials Testing Reactor, an increase in grain boundary P segregation associated with decrease in carbon segregation was observed with increasing fluence. Relative significant hardening was recognized for the material with high bulk P content. Shifts in the reference temperature T0 and T41J were comparable for the thermally aged and irradiated materials.

Copper and phosphorus effect on residual embrittlement of irradiated model alloys and RPV steels after annealing

The dependence of the recovery of the transition temperature shift after annealing (475 °C, 100 h) on copper and phosphorus contents has been studied on irradiated reactor pressure vessel (RPV) materials. A set of model alloys with low nickel content, lower than 0.2 mass%, was used for the study. Copper and phosphorus contents were varied in a wide range: 0.005–0.99 and 0.002–0.039 mass%, respectively. Recovery efficiency has been estimated by the value of residual embrittlement after annealing, measured in terms of a shift in transition temperature ( Δ T K res ). A comparison of the results obtained on model alloys with data for VVER-440 RPV materials has also been carried out. Comparative analysis has confirmed the conclusion that Δ T K res is independent of phosphorus content while the effect of copper on Δ T K res is not significant for typical VVER-440 RPV materials with a typical range of Cu contents between 0.10 and 0.24 mass%. However, for model alloys with a wider range of copper content, copper mainly controls the value of Δ T K res .

Nondestructive SANS Study of Reactor Pressure Vessel Steel after Irradiation

The irradiation induced defects of irradiated reactor pressure vessel(RPV) steel were investigated by a small angle neutron scattering. The degradation of the mechanical properties of RPV steels during an irradiation in a nuclear power plant is closely related to the irradiation induced defects. The size of these defects is known to be a few nanometers, and the small angle neutron scattering technique is regarded as the best non destructive technique to characterize the nano sized inhomogeneities in bulk samples. The RPV steel was irradiated in the HANARO reactor in KAERI. The small angle neutron scattering experiments were performed at SANS instrument in the HANARO reactor. Both unirradiated and irradiated RPV steels were measured and the SANS data of both steels were compared. The nano sized irradiation induced defects were quantitatively analyzed by SANS. The type of defects was also analyzed based on the SANS results, and the effect of the chemical composition of the RPV steel on the irradiation induced defects was discussed.

Influence of hydrogen on the toughness of irradiated reactor pressure vessel steels

The influence of hydrogen on the mechanical behaviour of different reactor pressure vessel (RPV) steels was investigated by tensile tests in correlation to the chemical composition, the neutron fluence, the hydrogen charging condition, the strain rate, and the temperature. Small-angle neutron scattering (SANS) experiments, hydrogen analyses and thermal desorption investigations were performed to prove the evidence of hydrogen trapping at irradiation defects. An increasing susceptibility to hydrogen embrittlement indicated by reduction of area was observed at room temperature with in situ hydrogen-charged specimens when loaded by low strain rates or with specimens which had been irradiated at low temperature. Generally, the susceptibility increases with increasing strength of the steels. At 250 °C hydrogen embrittlement was not evident. The results do neither prove that irradiation defects are favoured traps for hydrogen nor give evidence that hydrogen affects the RPV integrity under normal operating conditions.

Small-angle neutron scattering study of post-irradiation annealed neutron irradiated pressure vessel steels

Neutron irradiated reactor pressure vessel steels of two different Cu contents exposed to isochronal post-irradiation annealing treatments at stepwise increasing levels of temperature have been investigated. Results of small-angle neutron scattering experiments are reported and compared with the respective unirradiated and as-irradiated conditions. Volume distributions of scatterers and ratios of total-to-nuclear scattering intensities ( A-ratio) were calculated from the measured differential macroscopic scattering cross-sections. Finally, total volume fractions of scatterers, peak radii of the volume distribution and partial information on the composition of scatterers were extracted and discussed as a function of Cu level and annealing temperature. Our measurements indicate that irradiation-induced clusters dissolve completely for the lower Cu level of 0.15 wt% but dissolve only partly prior to the onset of coarsening for the higher Cu level of 0.29 wt%. An Arrhenius type behaviour and a remarkable reversibility of the irradiation-induced cluster distribution are observed in the complete dissolution regime.

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 ≈>10nm in the unirradiated and irradiated samples. However, the results do not show any evidence for hydrogen trapping in irradiation defects.

Multi-component clustering in VVER-type pressure vessel steels – thermodynamic aspects and impact on SANS

Atom probe field ion microscopic investigations of irradiated VVER 440-type reactor pressure vessel steels suggest the appearance of multi-component clusters. The effect is unexpected considering the phase diagrams of the concerning alloys. Numerical calculations of the negative minimum of the thermodynamic driving forces for a multi-component system are carried out considering a quasi-quaternary system consisting of Fe, Mn, Si and vacancies. A relative minimum was only found for a composite model of the multi-component clusters composed of an Fe containing core and a vacancies-rich shell. The ratio of their sizes is estimated from the condition that such structures agree with the small-angle neutron scattering (SANS) curves measured.

Dependence of Barkhausen noise in the neutron irradiated reactor pressure vessel steel

Dependence of Barkhausen noise in the neutron irradiated reactor pressure vessel steel

Dependence of Barkhausen noise in the neutron irradiated reactor pressure vessel steel

Barkhausen noise (BN) and ferromagnetic resonance (FMR) experiments were conducted in the neutron irradiated commercial nuclear pressure vessel steel. The energy of BN decreased by neutron irradiation and the rates increased with increasing neutron dose. The resonance field and linewidth determined from FMR spectra increased with increasing neutron dose. The experimental results were explained as domain wall pinning of irradiation induced precipitates and change of anisotropy energy associated with neutron irradiation.

The Study of Neutron Irradiated 15Ch2MFA Reactor Pressure Vessel Steel

The Study of Neutron Irradiated 15Ch2MFA Reactor Pressure Vessel Steel

Prediction of radiation hardening in reactor pressure vessel steel based on a theoretical model

A theoretical model is presented that calculates the amount of hardening in reactor pressure vessel (RPV) steels under neutron irradiation. A reaction rate theory is used to describe the evolution of radiation-induced microstructures. The point defect clusters (PDC) and copper-rich precipitates (CRP) have been recognized as contributing to radiation hardening in irradiated alloys, which act as barriers to the dislocation motion on the slip plane. The model includes a detailed description of PDC and CRP, and estimates the increase in yield strength due to these microstructures. The yield strength changes predicted by the model calculation are compared with the measured yield strengths, which were obtained from a surveillance program in French pressurized water reactors (PWR) and from post-irradiation tests performed at a high-flux advanced neutron application reactor (HANARO) in Korea. A fair agreement was found between estimated and observed hardening of irradiated RPV steels in spite of uncertainty about a broad range of irradiation and material parameters. The calculation results imply that the inclusion of CRP tends to overestimate the amount of radiation hardening in low-copper RPV steels. Therefore, the PDC evolution is more important for predicting the amount of radiation hardening in low-copper steels.

Application of Master Curve fracture toughness for reactor pressure vessel integrity assessment in the USA

The Master Curve fracture toughness approach has been used in the USA for better defining the transition temperature fracture toughness of irradiated reactor pressure vessel (RPV) steels for end-of-life (EOL) and EOL extension (EOLE) time periods. The first application was for the Kewaunee plant in which the life-limiting material was a circumferential weld metal. Fracture toughness testing of this weld metal corresponding to EOL and beyond EOLE was used to reassess the PTS screening value, RT PTS, and to develop new operating pressure–temperature curves. The NRC has approved this application using a shift-based methodology and higher safety margins than those proposed by the utility and its contractors. Beaver Valley Unit 1, a First Energy nuclear plant, has performed similar fracture toughness testing, but none of the testing has been conducted at EOL or EOLE at this time. Therefore, extrapolation of the life-limiting plate data to higher fluences is necessary, and the projections will be checked in the next decade by Master Curve fracture toughness testing of all of the Beaver Valley Unit 1 beltline materials (three plates and three welds) at fluences near or greater than EOLE. A supplemental surveillance capsule has been installed in the sister plant, Beaver Valley Unit 2, which has the capability of achieving a higher lead factor while operating under essentially the same environment. The Beaver Valley Unit 1 evaluation has been submitted to the NRC. This paper reviews the shift-based approach taken for the Beaver Valley Unit 1 RPV and presents the use of the RT T 0 methodology (which evolved out of the Master Curve testing and endorsed through two ASME Code Cases). The applied margin accounts for uncertainties in the various material parameters. Discussion of a direct measurement of RT T 0 approach, as originally submitted for the Kewaunee case, is also presented.

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.

Thermally activated deformation of irradiated reactor pressure vessel steel

Temperature and strain rate change tensile tests were performed on two VVER 1000-type reactor pressure vessel welds with different contents of nickel in unirradiated and irradiated conditions in order to determine the activation parameters of the contribution of the thermally activated deformation. There are no differences of the activation parameters in the unirradiated and the irradiated conditions as well as for the two different materials. This shows that irradiation hardening preferentially results from a friction hardening mechanism by long-range obstacles.

Computer simulation of the core structure of the screw dislocation in α-iron containing copper precipitates: II. dislocation–precipitate interaction and the strengthening effect

The small, coherent BCC precipitates of copper that form during fast-neutron irradiation of ferritic steels are an important component in the irradiation hardening that occurs during service. The conventional explanation of hardening due to copper precipitates is given in terms of the Russell–Brown (M. Acta Metall. 20 (1972) 969) modulus hardening model, in which precipitates are treated as soft spots in iron. In the present paper, the core structure and energy of a <111> screw dislocation are computed as it approaches a row of precipitates. The results indicate that the hardening observed in experiments is due to the effect of the screw dislocation core on the BCC copper structure rather than elastic interaction. This is a new precipitate strengthening effect. The increase in the flow stress is estimated from the interaction energy between the dislocation and the precipitate row, and the estimated value for precipitates of a size and spacing found in irradiated reactor pressure vessel steels is encouragingly close to that found experimentally.

Microstructural characterization of irradiation-induced Cu-enriched clusters in reactor pressure vessel steels

The effect of irradiation on microstructure of four irradiated reactor pressure vessel steels (a low copper A533B-1 plate, a low copper A508-3 forging, a high copper Linde 80 flux weld and a high copper Linde 1092 flux weld) was determined by using complementary microstructural techniques such as optical position-sensitive atom probe (OPoSAP), field emission gun scanning transmission electron microscopy (FEGSTEM) and small angle neutron scattering (SANS). In the low copper steels, irradiation resulted in small shifts in transition temperature and small changes in hardness increments. The microstructural analyzes showed that this response was dominated by matrix damage. In contrast, both copper-enriched clusters and matrix damage formed in the high copper welds. This information was then used as input to the Russell–Brown model to predict the change in hardness resulting from copper-enriched clusters. The calculated hardness increments were found to be consistent with the experimental data.

Application of the miniature specimen technique to material irradiation tests and surveillance for reactor components

Yield strength, tensile strength and uniform elongation of unirradiated and irradiated reactor pressure vessel steels were determined by the modified miniature specimen test (MMST). The facility and conditions for irradiating the miniature specimens are described. The results were compared with conventional tensile test data. Empirical formulas for correcting result errors caused by the deviation of specimen thickness were established. The application of the MMST technique to radiation surveillance for pressurized water reactor components is addressed.

Views of TAGSI on the principles underlying the assessment of the mechanical properties of irradiated ferritic steel Reactor Pressure Vessels

The principles are reviewed which underlie trend curve development for irradiated ferritic Reactor Pressure Vessel steels where the dominant embrittlement mechanism is cluster hardening. The extent to which the form of the equations employed reflect the underlying physical mechanism is assessed.