Shift Of Ductile-brittle Transition Temperature Research Articles

Numerical prediction of the effect of irradiation on the Charpy Upper Shelf Energy of Reactor Pressure Vessel steels

Neutron irradiation affects the mechanical properties of Reactor Pressure Vessel (RPV) steels used in Light Water Reactors (LWR). Fracture properties are particularly monitored based on surveillance programs, mostly through Charpy impact tests. The evolutions of Ductile Brittle Transition Temperature (DBTT) and Upper Shelf Energy (USE) with irradiation are evaluated and used for structural integrity assessment depending on regulation procedures. Semi-empirical correlations have been proposed to predict DBTT and USE shifts as a function of material chemical composition and fast neutron fluence, based on extended experimental database. In this study, numerical simulations of Charpy impact tests are performed in order to predict quantitatively the evolution of USE with irradiation. In a first part, the mechanical properties of an A508 Cl3 RPV steel are presented, including tensile tests on flat and notched samples and Charpy impact tests, for various irradiation levels and test temperatures. USE values are found to be within the bounds of correlation curves. In a second part, the experimental data are used to calibrate constitutive equations describing the material behavior up to high strain levels. Charpy impact test finite element simulations are then detailed, including ductile fracture modelling. In a third part, the ductile fracture model is calibrated based on the experimental data, and numerical simulations are used to predict the evolution of USE with irradiation. The decrease of USE with irradiation is recovered, and the numerical results are found to be in quantitative agreement with the correlation curves. Additional simulations confirm the agreement for different reference unirradiated USE and chemical composition, as well as using a simplified ductile fracture model. These results indicate that numerical simulations of Charpy impact test relying on calibrated constitutive equations can be used to predict USE and required experimental characterization are finally discussed.

Radiation embrittlement of clad-HAZ of RPV of a decommissioned BWR plant

Radiation hardening/embrittlement was investigated for clad-heat affected zone (clad-HAZ)) of reactor pressure vessel (RPV) of Hamaoka unit-1 boiling water reactor (BWR) under decommissioning process. The neutron irradiation fluence in the HAZ of the RPV steel was 1.5 × 1022 n/m2 (E > 1MeV). The shift in ductile-brittle transition temperature, ΔDBTT, was measured to be +26 °C using 1/3-sized Charpy specimens, while the increase in Vickers hardness, ΔHv, was 19 HV in the clad-HAZ of 1.6 mm depth from the clad boundary. Atom probe tomography (APT) was conducted at the locations 0.1, 0.8, 1.6. 4 and 10 mm away from the clad boundary, and revealed that solute clusters were formed in both clad-HAZ and base metal (10 mm depth), and the clusters were rich in Mn, Ni, Si, Cu with the averaged diameter of about 2.5 nm and the number density of 2 × 1022/m3, being almost independent of depth location. As for the chemical compositions of the clusters, Mn content increased with decreasing Cu content. This suggests that there might be site completion between Mn and Cu for clustering. Radiation hardening of clad-HAZ induced by the formation of solute clusters was estimated on the bases of the dispersed barrier hardening (DBH) model in terms of N·d, and well interpreted with α = 0.06–0.08. The ΔDBTT-Vf plot, of which the data was measured with 1/3 size specimens sampled from the clad-HAZ at 1.6 mm depth, agreed with the previous surveillance test results measured in surveillance tests with standard size specimens of base metal.

Effect of Microstructure on High-Strength Low-Alloy Steel Welded Joint Toughness with Simulation of Heat-Affected Zone Coarse-Grained Area

A simulated coarse-grained heat affected zone microstructure formation mechanism is established in high strength low alloy steels using electron backscatter diffraction (EBSD). The governing effect of dispersion and ratios between different types of ferritic structural constituents on variation in impact strength is demonstrated. It is assumed from results of simulating a heat affected zone coarse-grained area that a reduction in welding energy input leads to a shift in ductile-brittle transition temperature towards a lower temperature.

Grain-boundary phosphorus segregation in highly neutron-irradiated reactor pressure vessel steels and its effect on irradiation embrittlement

Reactor pressure vessel (RPV) steels for pressurized water reactors (PWRs) with bulk P contents ranging from 0.007 to 0.012wt.% were subjected to neutron irradiation at fluences ranging from 0.3 to 1.2 × 1020 n/cm2 (E > 1 MeV) in PWRs or a materials testing reactor (MTR). Grain-boundary P segregation, which was analyzed using Auger electron spectroscopy (AES) on intergranular facets, increased with increasing neutron fluence. A rate theory model based on four diffusion-reaction equations for substitutional P atoms, octahedral interstitial P atoms, vacancies, and self-interstitial atoms was also used to simulate the increase in grain-boundary P segregation for RPV steels with a bulk P content up to 0.020wt.%, using parameters optimized by the present AES data. The increase in grain-boundary P segregation in RPV steel with a bulk P content of 0.015wt.%, which is the maximum P concentration in RPV steels used in Japanese nuclear power plants intended for restart, was estimated to be less than 0.1 in monolayer coverage at 1 × 1020 n/cm2 (E > 1 MeV). A comparison of the PWR data with the MTR data, including that from the literature, showed that neutron flux had no effect upon grain-boundary P segregation for A533B steels. The relationships of the ductile-brittle transition temperature (DBTT) shifts to grain-boundary P segregation and to yield strength were also discussed. A linear relationship between the yield strength and the DBTT shift with a slope of 0.63 was obtained for RPV steels with a bulk P content up to 0.026wt.%, which is higher than that of most U.S. A533B steels. It is concluded that the intergranular embrittlement is unlikely to occur for RPV steels irradiated in PWRs.

Degradation of impact toughness in cast stainless steels during long-term thermal aging

Cast austenitic stainless steels (CASSs) have been extensively used for the large components of nuclear reactor primary coolant systems. Since the cast steels inevitably contain degradable metastable phases and replacement of the large coolant system components is impractical, the thermal embrittlement of CASS components has been a serious concern in the extended-term operation of nuclear power plants. This study aimed to systematically measure and analyze the effect of long-term thermal aging on the Charpy impact toughness to provide a comprehensive understanding of thermal degradation behavior and a practical aging model to predict the degree of thermal degradation in the cast stainless steels. The materials tested in the research include eight CASS alloys (two CF3s, one CF3M, three CF8s, and two CF8Ms) and two reference wrought materials (304L and 316L), in which the nominal δ-ferrite content ranges from ~2% to 33%. These stainless steels have been thermally aged at two light water reactor (LWR) temperatures (290 and 330 °C) and at two accelerated-aging temperatures (360 and 400 °C) for up to 30,000 h; these include both under-aged and over-aged conditions relative to the extended service lifetime (80 years). Charpy impact testing was performed for aged and non-aged specimens, and the impact (absorbed) energy parameters were correlated with a new aging parameter (A). Both the reduction of impact fracture toughness and the shift of ductile-brittle transition temperature were strongly dependent on the δ-ferrite content and degree of thermal aging. A linear relationship was found between the increasing rate of the index transition temperature T41J and aging parameter A; base on which an empirical model was proposed for prediction of the transition temperature as a simple function of the aging parameter (A) and δ-ferrite content (F). Finally, the critical aging parameter for embrittlement (AC) was evaluated and compared with the existing δ-ferrite content criteria.

Scientific basis for cold brittleness of structural BCC steels and their structural degradation at below zero temperatures

The paper considers the physics of cold brittleness of structural bcc steels and methods of reducing the ductile-brittle fracture temperature. A complex study was performed to examine the degradation of structural phase state of pipe steel 09Mn2Si from the main gas pipeline of Yakutia after long-term (over 3 0 years) operation. Important regularities of degradation of pearlite colonies with carbide precipitation on ferrite grain boundaries were revealed. This phenomenon is associated with brittle fracture of gas pipelines. It is shown that the low-temperature kinetic processes in main pipelines which define the degradation of their structure and properties are related to interstitial athermal structural states in the zones of local crystal structure curvature. This is a fundamentally new, as yet unknown, mechanism. Pipe steels in warm rolling acquire a longitudinal textured band structure with alternating bands of initial ferrite grains and bands of fine grains with carbide precipitates formed during lamellar pearlite degradation. This type of structure allows for a shift of ductile-brittle transition temperature down to -80°C and ductility δ = 22% at this temperature. The production of high-curvature vortex structure in pipe steel surface layers results in a 3.5-fold increase in their service life.

Fracture transitions in iron: Strain rate and environmental effects

Abstract

Analysis of a link of embrittlement mechanisms and neutron flux effect as applied to reactor pressure vessel materials of WWER

The effect of neutron flux on embrittlement of WWER RPV materials is analyzed for cases when different radiation defects prevail. Data bases on the ductile–brittle transition temperature shifts obtained in the surveillance specimens programs and the research programs are used. The material embrittlement mechanisms for which the flux effect is practically absent and for which the flux effect is remarkable are determined. For case when the phosphorus segregation mechanism dominates the theoretical justification of the absence of the flux effect is performed on the basis of the theory of radiation-enhanced diffusion.

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic–martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile–brittle transition temperature shifts. For neutron doses of 1.5–2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350–500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.

RADIATION-INDUCED EMBRITTLEMENT AND LIFE EVALUATION OF REACTOR PRESSURE VESSELS

Neutron irradiation can lead to the embrittlement and the shift of ductile-brittle transition temperature (DBTT) of the reactor pressure vessel (RPV) steels, which is the potential threat to the safe operation of reactors. Surveillance program for RPVs, irradiation damage and embrittlement of RPV steels and safe evaluation of RPVs were reviewed in this paper.

Solid-state diffusion bonding of high-Cr ODS ferritic steel

Solid-state diffusion bonding (SSDB) was employed to join high-Cr oxide dispersion strengthened (ODS) ferritic steel (Fe–15Cr–2W–0.2Ti–0.35Y 2O 3) blocks under uniaxial hydrostatic pressure using a high-vacuum hot press, and the microstructure and mechanical properties of the joints were investigated. High-Cr ODS ferritic steels were successfully diffusion bonded at 1200 °C for 1 h, without precipitates and microvoids at the bonding interface or degradation in the base materials. Transmission electron microscopic observation revealed that the nano-oxide particles near the bonding interface were uniformly distributed in the matrix and that the chemical composition across the bonding interface was virtually constant. At room temperature, the joint had nearly the same tensile properties and exhibited anisotropic behavior similar to that of the base material. The tensile strength of the joint region at elevated temperatures is nearly the same as that of the base material, with necking behavior at several micrometers from the bonding interface. The total elongation of the joint region decreased slightly at 700 °C, with an exfoliation fracture surface at the bonding interface. Although a small ductile–brittle transition temperature shift was observed in the joints, the upper shelf energy was fully reserved in the case of joints with L–R bonding orientation, for which cracks did not propagate on the bonding interface. Therefore, it is concluded that SSDB can be potentially employed as a joining method for high-Cr ODS ferritic steel owing to the excellent mechanical properties of the joints.

Combined effects of phosphorus segregation and partial intergranular fracture on the ductile–brittle transition temperature in structural alloy steels

This article introduced effective P segregation in terms of the product of the amount of segregated P and partial intergranular fracture to account for the combined effects on the ductile–brittle transition temperature (DBTT) measured by dynamic and static notched bar tests on several alloy steels. Effective P segregation characterized a DBTT shift caused by thermal ageing and/or neutron irradiation in various A533B and 2.25Cr–1Mo steels, while P segregation controlled the DBTT independent of the fraction of intergranular fracture in thermally aged 3.5Ni–1.7Cr steels with and without Mo and V. These DBTT behaviors of the two classes of steels are related to differing brittle fracture mechanisms. The present analysis enables one to investigate how the embrittling potency of segregated P in association with partial intergranular fracture is influenced by the differing plasticity and the addition of several alloying elements under dynamic and static loading conditions.

Irradiation-Induced Hardening and Embrittlement of High-Cr ODS Ferritic Steels

Abstract The dependence of irradiation temperature, neutron dose, and Cr content on irradiation hardening and embrittlement has been investigated for oxide dispersion strengthened (ODS) ferritic steels irradiated from 290 to 600°C up to the maximum neutron dose of 0.75 dpa. Irradiation hardening and embrittlement of the materials strongly depend on the irradiation temperature. While neutron dose increases from 0.01 to 0.21 dpa, the ODS steels with Al show a saturation of irradiation hardening but the no-Al ODS steel has an increase in the hardening. The higher the Cr content, the more significant the irradiation hardening and embrittlement of ODS steels irradiated below 420°C. Interestingly, the ODS ferritic steels exposed to neutrons show little change of tensile elongation, compared to 9Cr JLF-1. Furthermore the ductile-brittle transition temperature (DBTT) shift of ODS steels increases with the increasing irradiation hardening, indicating that the increase of yield strength due to radiation induced hardening increases the DBTT and thus a reduction of hardening is effective to reduce the DBTT shift.

Influence of TiN Inclusions on the Cleavage Fracture Behavior of Low-Carbon Microalloyed Steels

Toughness is a major concern for low-carbon microalloyed steels. In this work, the impact fracture behavior of two low-carbon Ti-V microalloyed steels was investigated in order to better understand the role of TiN inclusions in the toughness of the steels. The steels had similar chemical compositions and were manufactured by the same rolling process. However, there was an obvious difference in the ductile brittle transition temperature (DBTT) in the Charpy V-notch (CVN) impact tests of the two steels; one (steel 1) possessing a DBTT below −20 °C, while the DBTT of the other (steel 2) was above 15 °C. Scanning electron microscopy (SEM) fractography revealed that there were TiN inclusions at the cleavage fracture initiation sites on the fracture surfaces of steel 2 at both low and room temperatures. It is shown that the TiN inclusions had nucleated on Al2O3 particles and that they had pre-existing interior flaws. A high density of TiN inclusions was found in steel 2, but there was a much lower density in steel 1. Analysis indicates that these inclusions were responsible for the shift of DBTT to a higher temperature in steel 2. A mechanism is proposed for understanding the effect of the size and density of TiN inclusions on the fracture behavior, and the cleavage fracture initiation process is analyzed in terms of the distribution and development of stresses ahead of the notch tip during fracture at both low and room temperatures.

Small specimen test technique for evaluating fracture toughness of blanket structural materials

Fracture toughness of blanket structural materials is one of the important material engineering properties to design structural components and evaluate the degradation during reactor operation. The present study showed that the master curve (MC) methodology to evaluate change in fracture toughness in the cleavage regime of a reduced-activation ferritic (RAF) steel before and after a thermal embrittlement treatment was applied by using sub-sized compact tension (CT) specimens. The reference temperature of MC for the JLF-1LN steel was estimated as −52 °C by using 1/2CT specimens. The shift of the reference temperature obtained by the MC methodology was judged to be somewhat larger than that of the ductile–brittle transition temperature shift obtained by the Charpy impact tests.

Assessment of Neutron Irradiation-Induced Grain Boundary Embrittlement by Phosphorous Segregation in a Reactor Pressure Vessel Steel

Abstract The materials used were two sorts of reactor pressure vessel steels (RPVSs), which contain different amounts of impurity phosphorous (P) and copper (Cu). The specimens for Charpy V-notch impact tests and Auger electron spectroscopy (AES) were irradiated in the Japan Materials Test Reactor (JMTR) at 290°C up to fluences of 6 × 1021 and 1 × 1024 n/m2 using a multi-division temperature control capsule which enables removal of a part of the sub-capsules during operation of the reactor. Neutron irradiation resulted in a significant shift in the ductile-brittle transition temperature (DBTT) accompanied by a large irradiation hardening in the high P and high Cu steel. The AES measurements following the irradiations revealed that almost no phosphorous segregation occurred at grain boundaries. The DBTT shifts by neutron irradiation were reasonably interpreted in terms of a so-called hardening mechanism. The SEM observations of the fractured surface indicated that a very small amount of grain boundary fracture was induced at the irradiation conditions, and resultantly no grain boundary embrittlement was observed. Based on the results of irradiation experiments as well as long-term thermal aging experiments beyond 16 000 h, neutron irradiation-induced grain boundary embrittlement is considered to rarely happen to the RPVS that contains 110 wppm of phosphorous.

Effects of Nickel Addition on Microstructural Evolution and Mechanical Properties of Reduced Activation Martensitic Steels Irradiated in the ATR-A1

Validity of nickel isotope tailoring (NIT) method to be able to generate a large amount of transmutation helium in 9%Cr-2%W reduced activation ferritic (RAF) steel for fusion reactor structural material was investigated and discussed through mechanical properties and microstructural evolution after fission neutron irradiation. Miniature tensile and Charpy impact specimens of RAF steels and 1% nickel added RAF steel were irradiated in the ATR-A1 to evaluate irradiation hardening and shift in ductile-brittle transition temperature (� DBTT). The amount of transmutation helium in all the RAF steels with and without nickel addition was calculated as only about 0.6 appm. After irradiation at 621 K, theDBTT for the 1% nickel added steel was similar for the JLF-1. After irradiation at 543 K, however, theDBTT for the 1% nickel added steel was significantly larger than that for the JLF-1. Microstructure observations revealed that irradiation-induced dislocation loops in the 1% nickel added steel were finer and denser than in the RAF steel without nickel addition, suggesting that the nickel addition to the RAF steels directly affected nucleation and growth processes of dislocation loops and enhanced irradiation hardening and embrittlement. Therefore, there is a limit in the NIT method as a simulation method for understanding effects of helium on irradiation embrittlement of RAF steels. The RAF steels used in the present study were JLF-1, Mod.JLF-1/LSM and Mod.JLF-1/LSM/Ni. (The ''LSM'' means low concentration of Si and Mn compared with the JLF-1.) The chemical compositions and the heat treatments of these steels are shown in Table 1. Mod.JLF-1/LSM/Ni contained 1 weight percentage of natural nickel. Neutron irradiation was carried out in the two sub-capsules of the ATR-A1, AS4 and AS16, in the Advanced Test Reactor (ATR) at Idaho National Engineering and Environ- mental Laboratory for 132.9 Effective Full Power Days (EFPDs). 11) Irradiation temperature and displacement dam- age were estimated for the two sub-capsules as at 543 K to 2.2 dpa and 621 K to 3.4 dpa. These sub-capsules were equipped with gadolinium shields of 1.7 mm in thickness to reduce the thermal neutron flux in order to suppress trans- mutation effects. The amount of transmutation helium was also suppressed due to the thermal shields. Greenwood et al. calculated the concentration of transmutation helium as 0.6 appm in Mod.JLF-1/LSM/Ni and 0.5 appm in the other RAF steels without nickel addition, respectively. 12)

BおよびB+N添加した8Cr-2W(F82H)鋼の強度特性と微細組織に及ぼす熱処理効果

Effect of heat treatment on mechanical properties and microstructures of Fe-8Cr-2W-0.1C-0.2V-0.04Ta martensitic steel F82H doped with about 60 mass ppm B or both of 60 mass ppm B and 200 mass ppm N has been examined. The normalization was heated at temperatures from 950 to 1250°C for 1.8 ks, followed by air cooling or water quenching. After tempering treatment at 780°C or 750°C, the distributions of boron, boron nitride and oxygen were measured by a secondary ion mass spectrometry (SIMS). Optical microstructural observation and tensile and Charpy impact tests were performed also. In the boron doped F82H the tensile properties were similar to the non-doped F82H, but the ductile-brittle transition temperature (DBTT) shifted from -43°C to 15°C. SIMS images with high intensity of boron were observed in localized regions of the boron doped F82H. Water quenching reduced the DBTT shift, about 30°C, and the localized boron intensity was slightly decreased. In the boron and nitrogen doped tempered-F82H heat-treated by the water quenching from the normalizing temperature, the properties of tensile and Charpy impact were similar to the non doped F82H, and no pronounced localized boron image was observed in the SIMS image and no intensities of oxides and boron nitride were observed either.

Microstructure property analysis of HFIR-irradiated reduced-activation ferritic/martensitic steels

The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile–brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr–2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr–2WVTa).

Influence of acrylonitrile–butadiene–styrene (ABS) morphology and poly(styrene‐co‐acrylonitrile) (SAN) content on fracture behavior of ABS/SAN blends

AbstractThis study attempted to correlate morphological changes and physical properties for a high rubber content acrylonitrile–butadiene–styrene (ABS) and its diluted blends with a poly(styrene‐co‐acrylonitrile) (SAN) copolymer. The results showed a close relationship between rubber content and fracture toughness for the blends. The change of morphology in ABS/SAN blends explains in part some deviations in fracture behavior observed in ductile–brittle transition temperature shifts. © 2004 Wiley Periodicals, Inc. J Appl Polym Sci 92: 2606–2611, 2004