Low-alloy Reactor Pressure Vessel Research Articles

Comparison of PM-HIP to forged SA508 pressure vessel steel under high-dose neutron irradiation

Powder metallurgy with hot isostatic pressing (PM-HIP) is an advanced manufacturing process that is envisioned to replace forging for heavy nuclear components, including the reactor pressure vessel (RPV). But PM-HIP products must at least demonstrate comparable irradiation tolerance than forgings in order to be qualified for nuclear applications. The objective of this study is to directly compare PM-HIP to forged SA508 Grade 3 Class 1 low-alloy RPV steel at two neutron irradiation conditions: ∼0.5–1.0 displacements per atom (dpa) at ∼270 °C and ∼370 °C. PM-HIP SA508 experiences greater irradiation hardening and embrittlement (total elongation) than forged SA508. However, uniform elongation and approximate toughness are comparable across all irradiated materials, suggesting irradiated PM-HIP SA508 exhibits superior ductility at maximum load-bearing capacity. The irradiation hardening mechanism is linked to composition rather than fabrication method. Since PM-HIP SA508 has higher Mn and Ni concentration, it is more susceptible to irradiation-induced nucleation of Mn-Ni-Si-P (MNSP) nanoprecipitates and dislocation loops, which both contribute to hardening. Conversely, the forged material nucleates fewer MNSPs, causing dislocation loops to control irradiation hardening. These results show promise for the irradiation performance of PM-HIP SA508 and can motivate future nuclear code qualification of PM-HIP fabrication for RPVs.

Effect of material inhomogeneity on constraint loss measured with subsized C(T) fracture specimens of a reactor pressure vessel steel

In this study, the fracture behavior of the low-alloy reactor pressure vessel JRQ steel was investigated in the ductile-to-brittle transition region with two different subsized compact tension fracture specimens: 0.18T C(T) and 0.5T C(T). The specimens were extracted from two thin plates (surface and middle) of the large JRQ steel plate, which is well-known to be macroscopically inhomogeneous in terms of brittleness. In particular, the reference temperature T0 as defined in the ASTM-E1921 standard was calculated for the two plates with the 0.18T C(T) and 0.5T C(T) specimens. A significant difference in the reference temperature T0 of more than 90 °C as determined with the 0.5T C(T) data was found between the two plates indicating a clear macroscopic inhomogeneity between the two depths. The measured toughness of 0.18T C(T) specimens and corresponding T0 evaluation showed that the effect of constraint loss was more pronounced for the middle plate due to its lower strength and that the constraint loss effect can be confounded with inhomogeneity. A local approach to fracture model was applied to account for the constraint loss by calibrating a criterion based on critical stressed volume V* and critical stress σ* obtained by finite element simulations. A big difference in the local fracture stress σ* between the two plates was found and based on the deduced local criterion it was shown that the measured 0.18T toughness can be effectively predicted within the temperature range of the ductile-to-brittle transition.

Impact of Chloride on the Environmentally-Assisted Crack Initiation Behaviour of Low-Alloy Steel under Boiling Water Reactor Conditions

Low-alloy reactor pressure vessel steels have a rather low susceptibility to stress corrosion cracking (SCC) in a boiling water reactor (BWR) environment if the high-temperature water contains no anionic impurities. Recent investigations revealed that under oxidizing BWR normal water chemistry (NWC) conditions extremely small amounts of chloride, can cause very high SCC growth rates in these materials. Therefore, the effect of continuous and temporary chloride additions on the crack initiation behaviour was explored by a series of constant extension rate tensile (CERT) and constant load tests in high-temperature water. In an NWC environment, containing ≥2 ppb of chloride, strain-induced corrosion cracking (SICC) initiation occurred briefly after the onset of plastic yielding and at much smaller strains than in high-purity water. On the other hand, under reducing hydrogen water chemistry conditions with up to 700 ppb chloride, no SICC was detected up to very high strains. CERT experiments, with moderate short-term chloride transients before and during the loading, showed that even serious mechanical loading transients, one day after returning to high-purity water, did not result in early SICC initiation.

Study on hydrogen embrittlement and dynamic strain ageing on low-alloy reactor pressure vessel steels

• HE increased with the DSA susceptibility of RPV steels at 250−288 °C. • Concentrated hydrogen in high binding energy traps by straining in DSA regime. • Enhanced HE as a consequence of plasticity localization due to DSA and hydrogen. • HESIV and HELP are dominant mechanisms at 250−288 °C. Tensile tests in air with hydrogen pre-charged smooth specimens and slow strain rate tests with smooth and notched specimens in hydrogenated high-temperature water (HTW) at elevated temperatures (250−288 °C) on low-alloy reactor pressure vessel (RPV) steels revealed a softening in strength and a pronounced reduction in ductility, where the magnitude of hydrogen embrittlement (HE) increased with the dynamic strain ageing (DSA) susceptibility of the RPV steels. In hydrogen pre-charged specimens and in hydrogenated HTW, shear dominated transgranular fracture by microvoid coalescence with increasing amounts of macrovoids, quasi-cleavage regions and secondary cracking were observed. Thermal desorption spectroscopy showed an increase in the concentration of trapped hydrogen in high binding energy traps (vacancies & voids) induced by straining in DSA regime. The observed hydrogen effects on fracture behaviour is a consequence of plasticity localization resulting from the interaction between DSA and hydrogen. HESIV and HELP are the dominant HE mechanisms.

Hydrogen embrittlement on fracture resistance of low-alloy reactor pressure vessel steel with high dynamic strain aging at 288 °C

Fracture mechanics tests in hydrogenated high-temperature water at 288 °C on low-alloy reactor pressure vessel steel with high dynamic strain aging (DSA) susceptibility revealed reduced fracture resistance, accelerated crack growth and increased quasi-cleavage appearance than test in high-temperature air. Hydrogen-induced softening by HESIV and HELP concur in synergy with DSA at 288 °C, which was evidenced by crack-tip microstructural characterisations.

Cu nanoprecipitate morphologies and interfacial energy densities in bcc Fe from density functional theory (DFT)

We use quantum mechanical ab initio simulation to inform a model which predicts the structures of coherent Cu nanoprecipitates that are thought to form in low-alloy reactor pressure vessel (RPV) steels. Using density functional theory (DFT) we calculated the interfacial energy densities of {100}, {110}, {111}, {210}, {211} and {221} orientated Fe-Cu interfaces. These energy density values were used in an optimisation model to predict low energy Cu nanoprecipitate geometries for nominal radii ranging from 1 to 5 nm. Strain states parallel to the Fe-Cu interface were calculated using an embedded atom method (EAM) potential for similarly sized Cu nanoprecipitates. Fe-Cu interfacial energy densities under equivalent strains were calculated using DFT allowing comparison with the predictions of the EAM potential. The DFT calculations revealed the lowest energy Fe-Cu interface orientation to be the {110}. Accordingly, the geometry prediction optimisations found that regardless of size the predicted Cu nanoprecipitate surface geometries were dominated by the {110} orientation. Notably, as the Cu nanoprecipitates increased in size the ratio of the surface made up of non-{110} orientations was observed to proportionally increase. This allows the nanoprecipitate to take a more spherical form so reducing its total surface area. Additionally, the Fe-Cu interface strains predicted by the EAM potential for all Cu nanoprecipitate radii are sufficiently small that they do not significantly alter the calculated interfacial energy density values. This finding suggests interfacial strain does not play a significant role in determining the morphology of Cu nanoprecipitates.

Microstructural characterization of the synergic effects of dynamic strain ageing and hydrogen on fracture behaviour of low-alloy RPV steels in high-temperature water environments

Tensile tests in high-temperature air with pre-charged hydrogen and elastic plastic fracture mechanics tests in hydrogenated high-temperature water (HTW) at 250 and 288 °C on low-alloy reactor pressure vessel (RPV) steels revealed a clear but moderate reduction of ductility and fracture resistance, respectively. The observed behaviour is a consequence of synergistic effects between hydrogen embrittlement (HE) and the dynamic strain ageing (DSA), in which the HE was amplified by a high DSA susceptibility. The deformation microstructures in the vicinity of the crack tips in air and HTW of two RPV steels with high DSA susceptibility were characterized in detail. These investigations support the idea that the environmental degradation of fracture resistance in hydrogenated HTW was mainly due to the plasticity localization by the interaction between DSA and hydrogen in RPV steels. Synergistic effects of DSA and hydrogen lead to heterogeneous distribution of dislocations and formation of dislocation cells inside bainitic laths.

Evaluation of Quasi-Static and Dynamic Fracture Toughness on the Low-Alloy Reactor Pressure Vessel Steel JRQ in the Transition Region

Three point bending and impact tests with sub-sized Charpy specimens were performed on the JRQ reference steel for reactor pressure vessels. Quasi-static and dynamic fracture toughness data were calculated and the fracture behavior in the ductile to brittle transition region was evaluated within the frame of the master curve method (ASTM E1921). Specimens with shallow and deep cracks were studied and the respective influence of crack length and loading rate on the reference transition temperature was determined. The force-time curves of specimens with shallow cracks presented significantly smaller oscillations with respect to the absolute force, making the fracture toughness evaluation more accurate.

In-situ investigations of hydrogen influenced crack initiation and propagation under tensile and low cycle fatigue loadings in RPV steel

Present work aims to unveil the mechanism of hydrogen embrittlement (HE) in SA508 Grade 3 Class I low alloy reactor pressure vessel (RPV) steel. In-situ tensile and low cycle fatigue (LCF) tests are performed on specially designed specimens using tensile/fatigue testing stage under scanning electron microscope (SEM). Electrochemical hydrogen charging resulted in localized void formation at prior austenite grain boundaries (PAGBs) during tensile loading. Alongside the hydrogen induced weakening of PAGBs due to synergetic HELP (hydrogen enhanced localized plasticity) and HEDE (hydrogen enhanced decohesion) mechanisms of HE, fish-eyes formation around Al2O3–SiO2 type inclusions are the primary factors for hydrogen enhanced tensile properties degradation in subject RPV steel. During LCF loading, crack initiation and propagation is facilitated by long rod inter-lath cementite particles distributed along the bainitic ferrite lath boundaries in the un-charged specimen. In case of hydrogen charged specimen, the edge crack formed during LCF loading propagated through the specimen by cleavage. Predominantly plasticity (slip) driven transgranular crack propagation occurred in un-charged specimen. In contrary, hydrogen charging resulted in LCF crack to propagate in mixed intergranular and transgranular manner during early stages of propagation, whereas once the crack length exceeded 5 to 6 grains, cleavage type transgranular crack propagation was observed.

Environmental degradation of fracture resistance in high-temperature water environments of low-alloy reactor pressure vessel steels with high sulphur or phosphorus contents

The fracture behaviour of low-alloy reactor pressure vessel steels with various sulphur and phosphorus contents, different environmentally-assisted cracking (EAC) and temper embrittlement (TE) susceptibilities was evaluated by elastic plastic fracture mechanics tests in air and various simulated light water reactors environments. A moderate but clear reduction of fracture initiation resistance occurred in a) high-sulphur steel with high EAC susceptibility in oxygenated high-temperature water with aggressive occluded crevice environment and preceding EAC growth, and in b) high-phosphorus steel with high TE susceptibility, where the reduction of fracture resistance was most pronounced in hydrogenated high-temperature water.

Effect of dynamic strain ageing on environmental degradation of fracture resistance of low-alloy RPV steels in high-temperature water environments

Elastic plastic fracture mechanics tests on different low-alloy reactor pressure vessel (RPV) steels in air and various simulated light water reactor environments revealed a clear, but moderate reduction of the fracture initiation resistance of RPV steels with high susceptibility to dynamic strain ageing (DSA) in high-temperature water (HTW). In HTW, the reduction of fracture initiation resistance increased with decreasing loading rate and was most pronounced in hydrogenated HTW. The stronger reduction of fracture resistance in hydrogenated HTW was mainly due to the localization of plastic deformation by interaction between DSA and hydrogen.

Environmental Degradation Effect of High-Temperature Water and Hydrogen on the Fracture Behavior of Low-Alloy Reactor Pressure Vessel Steels

Structural integrity of reactor pressure vessel (RPV) in light water reactors (LWR) is of highest importance regarding operation safety and lifetime. The fracture behaviour of low-alloy RPV steels with different dynamic strain aging (DSA) & environmental assisted cracking (EAC) susceptibilities in simulated LWR environments was evaluated by elastic plastic fracture mechanics tests (EPFM) and by metallo- and fractographic post-test analysis. Exposure to high temperature water (HTW) environments at LWR temperatures revealed only moderated reductions in the fracture initiation and tearing resistance of low alloy RPV steels with high DSA or EAC susceptibility, accompanied with a moderate, but clear change in fracture morphology, which indicates the potential synergies of hydrogen/HTW embrittlement with DSA and EAC under suitable conditions. The most pronounced degradation effects occurred in a) RPV steels with high DSA susceptibility, where the fracture initiation and tearing resistance reduction increased with decreasing loading rate and were most pronounced in hydrogenated HTW and b) high sulphur steels with high EAC susceptibility in aggressive occluded crevice environment and with preceding fast EAC crack growth in oxygenated HTW. The moderate effects are due to the low hydrogen availability in HTW together with high density of fine-dispersed hydrogen traps in RPV steels. Stable ductile transgranular tearing by microvoid coalescence was the dominant failure mechanism in all environments with additional varying few % of secondary cracks, macrovoids and quasi-cleavage in HTW. The observed behavior suggests a combination of plastic strain localisation by the Hydrogen-enhanced Local Plasticity (HELP) mechanism, in synergy with DSA, and Hydrogen-enhanced Strain-induced Vacancies (HESIV) mechanism with additional minor contributions of Hydrogen-enhanced Decohesion Embrittlement (HEDE) mechanism.

On the Analysis of Clustering in an Irradiated Low Alloy Reactor Pressure Vessel Steel Weld.

Radiation induced clustering affects the mechanical properties, that is the ductile to brittle transition temperature (DBTT), of reactor pressure vessel (RPV) steel of nuclear power plants. The combination of low Cu and high Ni used in some RPV welds is known to further enhance the DBTT shift during long time operation. In this study, RPV weld samples containing 0.04 at% Cu and 1.6 at% Ni were irradiated to 2.0 and 6.4×1023 n/m2 in the Halden test reactor. Atom probe tomography (APT) was applied to study clustering of Ni, Mn, Si, and Cu. As the clusters are in the nanometer-range, APT is a very suitable technique for this type of study. From APT analyses information about size distribution, number density, and composition of the clusters can be obtained. However, the quantification of these attributes is not trivial. The maximum separation method (MSM) has been used to characterize the clusters and a detailed study about the influence of the choice of MSM cluster parameters, primarily on the cluster number density, has been undertaken.

Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen.

The influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low-alloy steels under simulated boiling water reactor conditions

The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy reactor pressure vessel steels was evaluated under simulated boiling water reactor conditions. In normal water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity water values within few 100h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen water chemistry environment is much higher.

The influence of ppb levels of chloride impurities on the strain-induced corrosion cracking and corrosion fatigue crack growth behavior of low-alloy steels under simulated boiling water reactor conditions

The effect of chloride on the strain-induced corrosion cracking (SICC) and corrosion fatigue (CF) crack growth behaviour in low-alloy reactor pressure vessel steels was evaluated under simulated boiling water reactor normal water chemistry conditions by slow rising load and cyclic constant load amplitude tests with air fatigue pre-cracked fracture mechanics specimens. Chloride in the ppb level range increased the SICC initiation susceptibility, but had almost no effect on the subsequent SICC and CF crack growth. A strong effect of chloride addition of 100ppb on CF crack growth was observed at intermediate corrosion potentials and very low loading frequencies only.

Effect of hydrogen on tensile behavior of low alloy steel in the regime of dynamic strain ageing

Low alloy steels typically used for reactor pressure vessel (RPV) in light water reactors may undergo different degradations and ageing mechanisms during service like fatigue, strain-induced corrosion cracking and corrosion fatigue or irradiation embrittlement, the latter being recognized as life limiting factor. There is growing concern that hydrogen, absorbed from the high temperature water environment and corrosion reactions, may potentially reduce toughness of RPV steels in synergy (or competition) with other embrittlement mechanisms like irradiation embrittlement, thermal ageing or dynamic strain aging (DSA). Strain rate, temperature and occurrence of DSA in these steels may affect the severity of the effect of hydrogen on toughness. The present investigation was envisaged to characterize the effect of hydrogen on tensile and fracture behavior of low alloy RPV steels at different strain rates and temperatures with a special emphasis on the synergy between DSA and hydrogen embrittlement. For this reason, tensile tests were carried out with as-received and hydrogen pre-charged specimens between 25 and 400 °C and at strain rates between 10-1 and 10-6 s-1. The fracture mode was evaluated by detailed post-test fractography in the scanning electron microscope. DSA in these steels was established by the occurrence of serrations, negative strain-rate sensitivity and a maximum/minimum in strength/ductility at intermediate temperatures and strain rates. The DSA peak and range were found to be shifted to lower temperatures with decreasing strain rates and vice-versa. The hydrogen pre-charging resulted in marginal softening and strain-rate dependent reduction in ductility at 250/288 °C. The hydrogen embrittlement and reduction in ductility were more pronounced and the strain rate range for hydrogen embrittlement significantly extended in the RPV steel with higher DSA susceptibility demonstrating some synergy between DSA and hydrogen effects, probably due to the localization of plastic deformation. In presence of hydrogen, shear dominated ductile fracture (microvoid coalescence) with varying amounts of quasi-cleavage regions and secondary cracking along the prior austenite grain boundaries were observed. A detailed investigation on these aspects and a tentative mechanistic explanation is presented in this paper.

Stress Corrosion Cracking Behavior in the Transition Region of Alloy 182/Low-Alloy Reactor Pressure Vessel Steel Dissimilar Metal Weld Joints in Light Water Reactor Environments

The stress corrosion cracking (SCC) behavior perpendicular to the fusion line in the transition region between the alloy 182 nickel-base weld metal and the adjacent low-alloy reactor pressure vessel (RPV) steel of simulated dissimilar metal weld joints was investigated under boiling-water reactor normal-water chemistry conditions at different stress intensities and chloride concentrations. A special emphasis was placed on the question whether a fast growing interdendritic SCC crack in the highly susceptible alloy 182 weld metal can easily cross the fusion line and significantly propagate into the adjacent low-alloy RPV steel. Cessation of interdendritic stress corrosion crack growth was observed in high-purity or sulfate-containing oxygenated water under periodical partial unloading or constant loading conditions with stress intensity factors below 60 MPa·m1/2 for those parts of the crack front that reached the fusion line. In chloride containing water, on the other hand, the interdendritic stress corrosi...

Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

The critical cleavage fracture stress of SA508 Gr.4N and SA508 Gr.3 low alloy reactor pressure vessel (RPV) steels was studied through the combination of experiments and finite element method (FEM) analysis. The results showed that the value of the local cleavage fracture stress, σF, of SA508 Gr.4N steel was significantly higher than that of SA508 Gr.3 steel. Detailed microstructural analysis was carried out using FEGSEM which revealed much smaller grains, finer and more homogenous carbide particles formed in SA508 Gr.4N steel. Compared with the SA508 Gr.3 steel currently used in the nuclear industry, the SA508 Gr.4N steel possesses higher strength and notch toughness as well as improved cleavage fracture behavior, and is considered a better candidate RPV steel for the next generation nuclear reactors.

Strain-induced corrosion cracking behaviour of low-alloy steels under boiling water reactor conditions

The strain-induced corrosion cracking (SICC) behaviour of different low-alloy reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated boiling water reactor (BWR)/normal water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP ⩾+50 mV SHE, ⩾0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness <350 HV5) and increased (once initiated) with increasing loading rates and with increasing temperature with a possible maximum/plateau at 250 °C. A minimum K I value of 25 MPa m 1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate d K I /d t, but were not dependent on the actual K I values up to 60 MPa m 1/2. A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200–250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.