Pressure Vessel Steels Research Articles

Radiation Damage of Reactor Pressure Vessel Steels Studied by Positron Annihilation Spectroscopy—A Review

Safe and long term operation of nuclear reactors is one of the most discussed challenges in nuclear power engineering. The radiation degradation of nuclear design materials limits the operational lifetime of all nuclear installations or at least decreases its safety margin. This paper is a review of experimental PALS/PLEPS studies of different nuclear reactor pressure vessel (RPV) steels investigated over last twenty years in our laboratories. Positron annihilation lifetime spectroscopy (PALS) via its characteristics (lifetimes of positrons and their intensities) provides useful information about type and density of radiation induced defects. The new results obtained on neutron-irradiated and hydrogen ions implanted German steels were compared to those from the previous studies with the aim to evaluate different processes (neutron flux/fluence, thermal treatment or content of selected alloying elements) to the microstructural changes of neutron irradiated RPV steel specimens. The possibility of substitution of neutron treatment (connected to new defects creation) via hydrogen ions implantation was analyzed as well. The same materials exposed to comparable displacement damage (dpa) introduced by neutrons and accelerated hydrogen ions shown that in the results interpretation the effect of hydrogen as a vacancy-stabilizing gas must be considered, too. This approach could contribute to future studies of nuclear fission/fusion design steels treated by high levels of neutron irradiation.

Investigation of microstructure and mechanical properties of AISI2205/DIN-P355GH steel joint by submerged arc welding

• In this study, AISI 2205 duplex stainless steel materials and P355GH pressure vessel steels are joined by using the same additional metal and powder, with the same parameters and submerged arc welding method. • Welded with SAW method, these materials were tested by eye, magnetic particle and radiographic examination methods to detect surface and sub-surface defects. No pores, voids or cracks were found in the materials after the magnetic particle test and radiographic test. In addition, solid spot and fixation of the welded samples worked, and welded samples showed no unacceptable angular distortion and high-low defect error. • Tensile, bending, impact notch and hardness tests were applied to determine the mechanical properties of welded samples. When the tensile test results are examined, it is seen that although the highest tensile strength of the main materials is in AISI2205 material, % elongation values are very close to each other. • Successful results have been achieved from all SAW of duplex stainless steel and pressure vessel steels. This result shows that additional metal, welding flux and welding parameters are suitable for joining duplex stainless steel and pressure vessel steels. In this study, duplex stainless (AISI2205) steel and pressure vessel (P355GH) steel have been joined by submerged arc welding method (SAW). Three different joining operations (AISI2205-AISI2205), (P355GH-P355GH) and (AISI2205-P355GH) were performed respectively. ELOX UP 309 L submerged arc welding wire and Eliflux Bss-A submerged arc welding powder were used for joining materials. Welded materials have been tested with eye, magnetic particle and radiographic examination methods were used to detect surface and subsurface defects. Tensile, bending, Charpy impact notch and hardness tests were then performed to determine the mechanical properties of welded samples. Additionally, microstructure investigations were made to determine the structural changes in the welding zone. Various NDE technqiues were adopted to investigate the quality of the welds. As a result of tensile tests, the rupture was made of welding metal in AISI2205-AISI2205 joints and P355GH in other joints was made of base material. No visible errors were observed in the welding area after bending test. When the V-notch impact energies of the materials were compared, the highest toughness value was measured in the ITAB of AISI2205 material and the lowest toughness value was measured in the P355GH weld metal. In hardness measurements, the highest hardness value was obtained in the HAZ of the AISI2205 material, whereas the lowest hardness was measured in the P355GH base material. Microstructure studies indicated that the weld metal comprised a dendritic structure, and the grains were observed to become coarse in the region adjacent to the melting boundary on the main material side.

Effect of Peak Temperature on Microstructure and Mechanical Properties of Thermally Simulated Welding Heat-Affected Zones for 09MnNiDR Steel

The microstructures and mechanical properties of various heat-affected zones (HAZs) for 09MnNiDR pressure vessel steel are systematically analyzed. The results show that at − 70 °C, the thermally simulated subcritical heat-affected zone (SCHAZ) reaches the impact toughness of 270 J, which is the highest among various HAZs. Owing to the appearance of martensite–austenite (M–A) constituents distributed along the grain boundary in the thermally simulated critical heat-affected zone and fine-grained heat-affected zone, the impact toughness sharply decreases compared with that of the SCHAZ. The impact toughness of the coarse-grained heat-affected zone (CGHAZ) reaches 20 J, which makes the CGHAZ the weakest zone. The microstructure of the CGHAZ is composed of relatively coarse bainite and ferrite. Moreover, the proportion of high-angle grain boundaries (HAGBs) in CGHAZ is the lowest, at only 22%. Coarse grain size, microstructural transformation, and different proportions of HAGBs significantly affect the impact toughness in each zone of HAZs.

Experimental study of the relationship between fracture initiation toughness and brittle crack arrest toughness predicted from small-scale testing

It is vital to prevent brittle cracks in large structures. This is particularly important for a number of industry sectors including offshore wind, Oil & Gas, and shipbuilding where structural failure risks loss of human life and loss of expensive assets. Some modern steels exhibit high Charpy energy – i.e. high initiation fracture toughness, but poor resistance to crack propagation – i.e. low crack arrest toughness. The correlation between initiation and arrest toughness measured through small-scale testing is investigated in five different steels, which include S355 structural steel (with two different thicknesses), X65 pipeline steel, two high strength reactor pressure vessel steels and EH47 shipbuilding steel. Small scale mechanical tests were carried out to characterise the materials’ properties and were compared to the materials’ microstructures. A wide range of tests were carried out, including instrumented Charpy, drop weight Pellini, fracture toughness, tensile testing, and optical microscopy. Nil ductility transition temperature (NDTT) is used to characterise a material’s arrest properties. Initiation fracture toughness correlated with higher upper shelf Charpy energy and smaller average grain sizes, as expected, however none of these correlated well with the arrest toughness measured through NDTT. The NDTT correlated most strongly with the T27J temperature which indicates the start of lower shelf of the Charpy curve. This correlation held for all materials including those where the NDTT lies on the upper shelf of the Charpy curve. While initiation fracture toughness can be predicted through high Charpy toughness and operation temperatures on the upper shelf, crack arrest behaviour should be predicted from characteristics of the ductile to brittle transition temperature, for example by using the T4kN from instrumented Charpy tests or T27J.

In-situ measurements of stress during thermal shock in clad pressure vessel steel using synchrotron X-ray diffraction

Thermal shocks are an important incident in operation of a pressure vessel which can have a significant impact on the structural integrity of the vessel. Often experiments that consider the state of the vessel before and after the thermal shock are used to evaluate the effects of the thermal shock. The studies can be complemented by time-resolved numerical simulations, which may be validated against the final state of the vessel obtained experimentally, to infer the transient response of the material. The transient response is important as the material experiences the highest level of stress in a short period which can induce catastrophic failure. This paper reports time-resolved experimental quantification of strain in reactor pressure vessel material during thermal shock measured by in-situ synchrotron diffraction. Specimens were extracted from a plate of nuclear pressure vessel steel with a nickel alloy cladding deposited by overlay welding. The specimens, with and without cracks, were subjected to thermal loading by heating then rapidly quenching the cladding in cold water. Strains were measured during thermal loading at a point near the crack tip from which the stress state around the crack tip was calculated and compared with a transient finite element model of the experiment. It was found that the peak near-tip stress occurred within the first second after the onset of rapid cooling. It was demonstrated from experimental measurements that the peak stress intensity factor occurred during thermal shock, rather than under steady conditions before or after the thermal shock. It was shown that although the finite element simulation predicts the steady state condition of the material after thermal shock, its transient response dependents significantly on a number of inputs with high uncertainty, making its time-resolved results unreliable for high-fidelity integrity assessments.

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.

Observation of Mn-Ni-Si-rich features in thermally-aged model reactor pressure vessel steels

Atom probe tomography was used to characterise two low-Cu (< 0.04 at. %) model steels after exposure to long-term thermal ageing. Mn-Ni-Si-rich features were observed to form after as little as 20,731 h (∼ 2.4 years) of ageing. The composition of these features were compared to those predicted by thermodynamic models and the similarities and differences are discussed.

Effects of Alloying Elements (C, Mo) on Hydrogen Assisted Cracking Behaviors of A516-65 Steels in Sour Environments.

This study examined the effects of alloying elements (C, Mo) on hydrogen-induced cracking (HIC) and sulfide stress cracking (SSC) behaviors of A516-65 grade pressure vessel steel in sour environments. A range of experimental and analytical methods of HIC, SSC, electrochemical permeation, and immersion experiments were used. The steel with a higher C content had a larger fraction of banded pearlite, which acted as a reversible trap for hydrogen, and slower diffusion kinetics of hydrogen was obtained. In addition, a higher hardness in the mid-thickness regions of the steel, due to center segregation, resulted in easier HIC propagation. On the other hand, the steel with a higher Mo content showed more dispersed banded pearlite and a larger amount of irreversibly trapped hydrogen. Nevertheless, the addition of Mo to the steel can deteriorate the surface properties through localized pitting and the local detachment of corrosion products with uneven interfaces, increasing the vulnerability to SSC. The mechanistic reasons for the results are discussed, and a desirable alloy design for ensuring an enhanced resistance to hydrogen assisted cracking (HAC) is proposed.

FEM analysis of residual stress induced by repair welding in SUS304 stainless steel pipe butt-welded joint

Abstract Repair welding is widely used to repair defects such as fatigue crack and stress corrosion crack in metal structures for prolonging their service lives. However, the residual stress induced by repair welding potentially threatens the safety of the repaired weldments, especially stainless steel pipe structures and pressure vessels used in nuclear power plants. Based on the MSC. Marc software, a thermal elastic plastic finite element method with the consideration of full 3-D model, moving heat source and advanced material model was developed to simulate residual stress distribution induced by repair welding. In the current study, the effectiveness of the developed computational approach was verified by a two-pass thin-walled butt-welded SUS304 steel pipe mock-up at first. Then, the residual stress distributions in a medium thickness SUS304 steel pipe butt-welded joint before and after repair welding were investigated numerically. The simulation results show that the peak value of axial residual stress after repair welding was significantly increased, while the variation of the maximum value of hoop residual stress was limited. In addition, the effect of repair length on the feature of residual stress distribution was studied numerically. The numerical results indicate that the peak value of repair-induced residual stress decreases with the repair length. The main reason is that the longer the repair welding length is, the less local constraint will be. The residual stress distribution in the repair area at service temperature was also discussed in this study. The results obtained from the current study will be helpful in evaluating the structural integrity of repaired weldments.

Modeling and stress analysis of aluminium alloy based composite pressure vessel through ANSYS software

The traditional materials based pressure vessels are having more strength. Due to their high weight to strength ratio and corrosive properties, but it is not suitable in the aerospace and oil and gas factories. In industries are extensively required for pressure vessels, which will have low weight to strength ratio without affecting the strength. In recent years, most of the sectors replace conventional materials with aluminum matrix composite materials. On the other hand, Aluminum matrix composite (AMCs) materials with their higher specific strength and these characteristics will reduce the structure's weight. In this research paper, the AA6082 alloy based aluminum matrix composites have been prepared by stir casting technique to test their mechanical properties under different weight percentages of reinforcement. Various mechanical studies have been done, such as tensile, impact, flexural, and hardness. For the same geometrical parameters of the steel pressure vessel, FE Analysis of AMCs composite pressure vessel is carried out, and stresses for different internal pressures are determined. And the design is carried out in design software solid works and analyze in ANSYS workbench. Then the results of steel pressure vessel and composite pressure vessel are compared for stress results.

Evaluation of Transition Temperature in Reactor Pressure Vessel Steels 6using the Fracture Energy Transition Curve from a Small Punch Test

The small punch (SP) test is one of the small specimen test techniques, and standardization of the SP test method for evaluating the mechanical properties of metallic materials is in progress. In this study, the impact transition temperature of reactor pressure vessel steels (RPV) in nuclear power plants was estimated using the draft standard SP test method. The SP fracture energy (ESP) and normalized SP fracture energy (ENSP) of the RPV steels were evaluated at various temperatures, and their transition curves were derived and compared to the transition curve in the Charpy V notch (CVN) test. The SP transition region appeared at a much lower temperature range than that of the CVN owing to the size and notch effect. Ductile brittle transition temperature (DBTT) in the SP transition curve showed a linear relationship with DBTT and T41J in the CVN transition curve. The ductile to brittle transition behaviors of SP specimens were analyzed using fractographs and compared with the transition curves in ESP and ENSP. ENSP started to decrease at the temperature at which the SP ductile to brittle transition behavior occurred, and this means that the ENSP transition curves were in good agreement with transition behavior in the SP test. However, the ESP transition curves did not match transition behavior. Using DBTT in the ENSP transition curve is appropriate to estimate the CVNDBTT and T41J.

Detection of neutron irradiaton induced degradation of reactor steel by magnetic method

A novel nondestructive method, based on step by step measurement of minor magnetic hysteresis loops was used to determine the neutron irradiation induced degradation of nuclear reactor pressure vessel steel material. This method is called Magnetic Adaptive Testing (MAT), and two types of reactor steel material were investigated: base material type 22NiMoCr37 and weld material type 18MND5. Charpy geometry samples were irradiated by neutrons at the BR2 in SCK•CEN. The neutron irradiation generated structural changes in the material. As a result of our measurement, it was shown that this degradation could be detected by MAT measurements. Charpy impact tests were also done on the investigated samples and the results of these measurement were compared with the nondestructively determined magnetic quantities. It was shown that a closely linear correlation existed between the ductile-to-brittle transition temperature (DBTT) and MAT parameters.

Microstructure and impact toughness of 16MND5 reactor pressure vessel steel manufactured by electrical additive manufacturing

Electrical additive manufacturing can improve manufacturing efficiency and reduce the cost of 16MND5 reactor pressure vessel steel. Impact tests were conducted to compare the impact toughness of 16MND5 steels manufactured by the electrical additive manufacturing and conventional forging, respectively. It is found that the impact toughness of electrical additive manufacturing specimen was slightly higher than that of conventional forging specimen. The characterizations of microstructure show that there were large ferrites and carbides in electrical additive manufacturing specimen. The fracture mechanisms of electrical additive manufacturing specimen were that microvoids or microcracks were prone to nucleate at the large ferrite/bainite interface and large carbide/bainitic ferrite interface, where the stress concentration was high. In addition, the block size and high-angle grain boundaries played a vital role in hindering crack propagation of electrical additive manufacturing specimen, helping to improve the impact energy and leading to a low ductile–brittle transition temperature. The results suggest that the electrical additive manufacturing technology was an effective method to enhance the impact toughness of 16MND5 steel.

Characterization of fracture toughness based on yield stress and successful application to construct a lower-bound fracture toughness master curve

The fracture toughness KJc of ferritic steels in the ductile-to-brittle transition region is important for analyzing the structural integrity of structures subjected to large temperature variations. In this study, a σYS-based lower-bound (LB) KJc master curve (MC) was developed to overcome the failure of the ASTM E1921 MC to predict KJc above room temperature (RT). The characteristic KJc, was first identified after finding KJc for the load P5, which is the upper limit load for KIc defined in ASTM E399. It can be approximated as KL = 0.6σYS (W)1/2 for various materials and specimen types. Then, the characteristic σYS was selected to be σYSRT/σYS. Assuming that KJc > KLRT at RT, KLRT was assumed to give an LB, which is empirically true for materials tested in accordance with ASTM E1921. Finally, the MC was formulated as KJcLB = KLRT (σYSRT/σYS)5.5, by applying the critical distance scaling method. KJcLB was successfully a LB for 715 unirradiated KJc data points, including eight heats of reactor pressure vessel (RPV) steel and seven heats of non-RPV steel. This MC can be constructed without conducting fracture toughness tests if σYSRT is known, because a temperature-based σYS MC exists.

The dominant mechanisms for the formation of solute-rich clusters in low-Cu steels under irradiation

The formation of nano-sized, coherent, solute-rich clusters (NSRC) is known to be an important factor causing the degradation of the macroscopic properties of steels under irradiation. The mechanisms driving their formation are still debated. This work focuses on low-Cu reactor pressure vessel (RPV) steels, where solute species are generally not expected to precipitate. We rationalize the processes that take place at the nanometer scale under irradiation, relying on the latest theoretical and experimental evidence on atomic-level diffusion and transport processes. These are compiled in a new model, based on the object kinetic Monte Carlo (OKMC) technique. We evaluate the relevance of the underlying physical assumptions by applying the model to a large variety of irradiation experiments. Our model predictions are compared with new experimental data obtained with atom probe tomography and small angle neutron scattering, complemented with information from the literature. The results of this study reveal that the role of immobilized self-interstitial atoms (SIA) loops dominates the nucleation process of NSRC.

Metadynamic recrystallization behaviors of SA508Gr.4N reactor pressure vessel steel during hot compressive deformation

The two-pass isothermal hot compression method was used to study the effect of different thermal deformation conditions on static recrystallization behavior in Ni–Cr–Mo series SA508Gr.4N low alloy steel with interval holding time ranging from 1 to 300 s, temperature ranging from 950 to 1150 °C, strain rate ranging from 0.01 to 1 s−1, true strains ranging from 0.1 to 0.2, and initial austenite grain size ranging from 175 to 552 μm. It can be concluded that the static recrystallization volume fraction gradually increases with the increase in the deformation temperature, strain rate, strain and pass interval, and the decrease in the initial grain size, which is mainly due to the increase in the deformation energy storage and dislocations. Moreover, strain-induced grain boundary migration is the nucleation mechanism for static recrystallization of SA508Gr.4N low alloy steel. Based on the stress–strain curve, the predicted value obtained from the established static recrystallization kinetics model is in good consistence with the experimental value, and the static recrystallization thermal activation energy of SA508Gr.4N steel was calculated as 264,225.99 J/mol.

Abnormal influence of impurity element phosphorus on the hot ductility of SA508Gr.4N reactor pressure vessel steel

The effect of phosphorous on the hot-ductility of SA508Gr.4N reactor pressure vessel (RPV) steel has been studied. The investigation is done with the use of a Gleeble machine in conjunction with optical metallography, scanning electron microscopy, and transmission electron microscopy. It is observed that both the undoped and P-doped steels exhibit hot-ductility troughs between 750 and 900 °C. However, the trough for the P-doped steel is much shallower than that for the undoped steel. At around 850 °C, the reduction of area (RA) of the undoped steel is quite low (~50%), in comparison to the P-doped steel (~75%). Since there is no grain-boundary precipitation of ferrite and carbides in both undoped and P-doped samples over this temperature range, it can be inferred that the enhancement in hot-ductility for the P-doped steel is independent of these two factors. Grain-boundary microanalysis shows the presence of grain-boundary segregation of P in the tested P-doped samples, implying that the P-induced hot-ductility improvement of this RPV steel may arise from the suppression of grain-boundary sliding by P grain-boundary segregation. Undoubtedly, the present effect of P is unusual and the underlying mechanism for this effect has been briefly discussed. It needs to be underlined that a minor addition of P can play a positive role in the hot-ductility improvement of SA508Gr.4N RPV steel.

In-situ WB-STEM observation of dislocation loop behavior in reactor pressure vessel steel during post-irradiation annealing

To better understand the thermal stability of dislocation loops formed by long-term neutron irradiation in reactor pressure vessel (RPV) steels, in-situ scanning transmission electron microscopy (STEM) observation was performed for a surveillance test specimen of a European pressurized water reactor (PWR). The surveillance test specimen was neutron irradiated to a fluence of 8.2 × 1023 neutrons•m−2. A membrane sample from the surveillance test specimen was annealed at 673 K and 723 K for 30 min using a heating holder, and the same area was in-situ observed under STEM. After annealing, dislocation loops with Burgers vectors of ½ 〈111〉 and 〈100〉 were quantitatively examined. When annealing temperature increased from 673 K to 723 K, the number of dislocation loops decreases, whereas the size of them becomes larger. Correspondingly, the proportion of 〈100〉 dislocation loops changes from 27% to 45%. The ratio of 〈100〉 to ½ 〈111〉 loops increases with annealing temperature rising. The evolution process of dislocation loops during annealing at 723 K was in-situ observed and analyzed to shed light on the transformation mechanism of dislocation loops going from ½ 〈111〉 to 〈100〉. It is the first time to directly observe that two ½ 〈111〉 dislocation loops collide with each other and coalesce to form a 〈100〉 dislocation loop. Moreover, small ½ 〈111〉 dislocation loops could be absorbed by large 〈100〉 dislocation loops, whereas the Burgers vector of 〈100〉 loops remained unchanged. Dislocation decoration occurs during annealing due to the interaction between dislocations and loops. The dislocations decorated by loops are pretty stable during the continuous annealing process, which is well explained by molecular dynamics simulation.

CFD study of the PTS experiment in ROCOM test facility

With the aging of nuclear reactors, embrittlement of the reactor pressure vessel (RPV) steel, as a consequence of routine operations, is highly probable. To ensure operational integrity and safety, prediction and mitigation of compromising damage, brought on by pressurized thermal shock (PTS) following an emergency procedure, is of utmost importance. Computational fluid dynamics (CFD) codes can be employed to predict these events and have therefore been an acceptable method for such assessments. In this paper, CFD simulations of a density driven ECC state in the ROCOM facility are analyzed. Obtained numerical results are validated with the experimental measurements. Considerable attention is attributed to the boundary conditions and their influence, specifically outlet definitions, in order to determine and adequately replicate the non-active pumps in the facility. Consequent analyses focused on initial conditions as well as on the temporal discretization and inner iterations. Disparities due to different turbulent modelling approaches are investigated for standard RANS models. Based on observed trends for different cases, a definitive simulation setup has been established, results of which have been ultimately compared to the measurements.

Effects of Neutron, Ion and Proton Irradiation on Nano-Indentation Hardness of RPV Steels

The irradiation embrittlement damage of reactor pressure vessel (RPV) steel is one of its primary failure mechanisms. In this work, neutron, ion and proton irradiation experiments were carried on the same commercial RPV steels with the same irradiation fluence under the same temperature of 292°C. Then the nano-indentation hardness tests were performed on the RPV steel before and after irradiation. The results show that the irradiation hardening effects are observed by means of nano-indentation technique under the above three irradiations, and the hardening features are basically the same. While the max variation and increase rate are obviously different between those irradiations. It is found that the main reason of the above differences are caused by different energies of irradiation energetic particles, resulting in different types and quantities of defects. The conclusions in this paper are helpful to select and compare different irradiation experiments to the research of RPV steels irradiation embrittlement damage.