Pressure Vessel Materials Research Articles

Scale factor of standard and mini Charpy specimens from VVER-1000 RPV materials

A database has been developed to establish a correlation between standard and mini-Charpy specimens with the scope of tests of 38 sets of specimens. A methodical procedure has been suggested and implemented which permits a twofold reduction of the scatter of experimental data as compared with a similar approach to VVER-440 reactor pressure vessel (RPV) materials.

Neutron deep penetration through reactor pressure vessel and biological concrete shield of VVER-1000 Mock-Up in LR-0 reactor

Evaluation of the neutron fluence values in a reactor pressure vessel together with surveillance specimens programs from reactor pressure vessel materials are among the most important parts of in-service inspection programs that are necessary for realistic and reliable assessment of the RPV’s residual lifetime. The neutron fluence values are determined by means of calculation. These calculation results are accompanied by measurements of induced activity of the activation foil placed in capsules behind the reactor pressure vessel at selected locations. These results can be used for neutron flux density adjustment in the corresponding parts of the reactor system. Such information is not complete, because for determination of the radiation damage, the flux on the inner side of reactor pressure vessel has to be known. If the flux attenuation ratio is known, then the neutron flux density on the outer side can be easily recalculated for the inner side. These attenuation factors, applicable to the power reactors, can be determined experimentally in the VVER-1000 Mock-Up placed in the LR-0 reactor. The comparison of experimental results with calculated ones increases the reliability of their estimations. This paper aims to evaluate the measured neutron flux densities at various VVER-1000 Mock-Up positions by means of comparison with the Monte Carlo simulation. The effect of the various nuclear data libraries on the calculation results is also studied.

Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

Electrochemical Behaviors of Various Alloys in Ferric Sulfate Solution

A508III low alloy steels (LAS) are often used as reactor pressure vessel materials in nuclear power plants. LAS could be exposed to primary water of nuclear power plants when stainless steel or Ni-based alloy corrosion resistant layers failed or as a result of small leaks in flanged joints valves, fittings, etc. Primary water contains boric acid and lithium hydroxide. Under leaking conditions, the primary water could be concentrated due to evaporation and become aerated as the results of contacting with the environment, producing varying degrees of external corrosion. There could be corrosion product such as rust or ferric ions at the leaking sites. Previous work has conformed that the corrosion rates of A508III steel increased dramatically when ferric sulfate was added in boric acid solution. While in the ferric sulfate solution, adding boric acid had no significant effect on the electrochemical behavior and corrosion rates, suggesting that ferric ions promoted the subsequent corrosion process. A508III LAS contains about ~97 wt.% Fe, with Cr, Ni, Mn and other elements added to improve the corrosion resistance, mechanical properties and processing performance, etc. The present work investigated the electrochemical behavior of various alloys and their main alloying elements in 0.05 mol/L Fe2(SO4)3 solution at 25 oC. Pure iron, Cr, Ni, A508III LAS, 430 steel, 316L steel, and Alloy 690 were used as working electrodes. Open circuit potential (OCP), electrochemical impedance spectroscopy and anodic polarization curve were measured. OCP values of A508III steel and iron in 0.05 mol/L Fe2(SO4)3 solution were lower than -0.52 V(SCE). The OCP of A508III steel is a little higher than that of iron. OCP values of the other materials were stable at relatively higher values that range from 0.37 V(SCE) to 0.56 V(SCE), and the ascending order was that 316L steel, 430 steel, Cr, Alloy 690, and Ni. The anodic polarization curves for various materials in 0.05 mol/L Fe2(SO4)3 solution are shown in Figure 1. Anodic polarization curves for A508III and iron were nearly the same, and neither current peak nor dissolution/passivation transition could be observed. Active-passivation transition appeared and single current peak were observed on the curves for 316L and 430 steel. The anodic polarization curves for Cr, 690 alloy and Ni showed different trends at different potential ranges. The EIS Nyquist plots of various materials in 0.05 mol/L Fe2(SO4)3 solution are shown in Figure 2. The Nyquist plots of A508III steel and iron exhibited two capacitive loops at higher and intermediate frequencies, and one inductive loop at low frequencies. The second capacitive loop was smaller than the first which suggest that the protectiveness of surface film is not good. The second capacitive loop of iron was a little larger than that of A508III steel and this suggest that protectiveness of iron was slightly better. The Nyquist plots of Cr, 316L steel, 430 steel and Alloy 690 exhibited two capacitive loops. The second capacitive loop was far greater than the first one, suggesting that the high protectiveness of the surface films. The size of second capacitive loop decreased in the order as 316L steel, Alloy 690, 430 steel, and Cr. This suggests that when immersed in 0.05 mol/L Fe2(SO4)3 solution, the surface film protectiveness of Cr was good. The Nyquist plots of Ni exhibited two capacitive loops at high and intermediate frequencies, and Warburg characteristics at low frequencies. The first capacitive loops were nearly the same for all materials in this context. The second capacitive loop of Ni was significantly greater than those of A508III steel and iron, but far smaller than those of 316L steel, Alloy 690, 430 steel and Cr. The amounts of Cr and Ni in A508III LAS were not enough to significantly affect the electrochemical behavior in the Fe2(SO4)3 solution. The electrochemical behaviors of A508III steel and iron were nearly the same. Protective films could form on the electrode surfaces at OCPs in 0.05 mol/L Fe2(SO4)3 solution if there was a certain Cr content in alloys. The reactivity of materials with certain Cr element was higher than that with certain nickel element under anodic polarization. Figure 1

Effects of aging time and temperature of Fe-1wt.%Cu on magnetic Barkhausen noise and FORC

Magnetic Barkhausen noise (MBN), hysteresis measurements, first order reversal curves (FORC), Vickers microhardness, and Transmission Electron Microscopy (TEM) analyses were performed on Fe-1wt.%Cu (Fe-Cu) samples isothermally aged at 700°C for 0.5 – 25 hours to obtain samples with different sized Cu precipitates and dislocation structures. Fe-Cu is used to simulate the thermal and irradiation-induced defects in copper-containing nuclear reactor materials such as cooling system pipes and pressure vessel materials. The sample series showed an initial increase followed by a decrease in hardness and coercivity with aging time, which is explained by Cu precipitates formation and growth as observed by TEM measurements. Further, the MBN envelope showed a continuous decrease in its magnitude and the appearance of a second peak with aging. Also, FORC diagrams showed multiple peaks whose intensity and location changed for different aging time. The changes in FORC diagrams are attributed to combined changes of the magnetic behavior due to Cu precipitate characteristics and dislocation structure. A second series of samples aged at 850°C, which is above the solid solution temperature of Fe-Cu, was studied to isolate the effects of dislocations. These samples showed a continuous decrease in MBN amplitude with aging time although the coercivity and hardness did not change significantly. The decrease of MBN amplitude and the appearance of the second MBN envelope peak are attributed to the changes in dislocation density and structure. This study shows that the effect of dislocations on MBN and FORC of Fe-Cu materials can vary significantly and should be considered in interpreting magnetic signatures.

Stress Analysis of Thick Spherical Pressure Vessel Composed of Transversely Isotropic Functionally Graded Incompressible Hyperelastic Materials

In this paper closed form analytical solution for stress components of thick spherical shell made of transversely isotropic functionally graded hyperelastic material subjected to internal and external pressure is presented. Reinforced neo-Hookean strain energy function with variable material parameters is used to model pressure vessel material. The material constants of strain energy function are graded along the radial direction based on a power law function and have been calculated from experimental data by using Levenberg-Marquardt nonlinear regression method. Stress components and stretches of pressure vessel have been obtained for centrally symmetric condition. Following this, profiles of extension ratio, deformed radius of sphere, normalized radial stress and normalized circumferential stress are plotted as a function of radius of sphere in the undeformed configuration for different material inhomogeneity parameter (m). The obtained results show that the inhomogeneity properties of FGMs structure parameter have a significant influence on the displacement, stretch and stresses distribution along the radial direction.

Analysis of the impact of position in fatigue cracks on the fracture toughness of thick-walled pressure vessel material

The testing of fracture toughness thick-walled pressure vessel material showed a vast range of different results. Pressure vessels were made by hot plastic deformation treatment. Three sets of specimens were cut-out from the vessel, each having a same orientation of fatigue cracks in relation to the vessel axis. Fracture toughness tests were performed on the SENB specimens in accordance with ASTM 1820-11. The results showed a large range of different results, even within each sets of specimens. The results of fracture toughness in the initiation zone of stable crack growth were processed by means of Weibull statistical analysis. Based on these results of a direction distribution coefficient m was obtained and results reliability interval of fracture toughness was defined.

Material selection for the gas containment system of a compressed natural gas carrier fleet

In this work, the possible exploitation of fiber-reinforced composites in the context of maritime transportation of compressed natural gas (CNG) is investigated. In addition to a more conventional steel configuration, two different fiber materials, carbon and glass, are considered as construction materials for pressure vessels (PVs) to be stored on board ships, with thickness optimized by FEM analysis.The considered scenario is represented by the transportation of CNG from an offshore well to a terminal on shore. Fleets of ships carrying CNG in pressure vessels manufactured with the investigated materials are generated by means of a ship synthesis model (SSM) software and compared on the basis of technical and economical indicators.The choice of the construction material influences considerably the weight of the PVs, which represent a major item of total ship weight and reflects directly on the general transport performances in terms of resistance, seakeeping and reliability in the service. On the other hand, capital as well as operating expenditures are considerably affected by the choice. When exploring the design space, the ship synthesis model is able, at a preliminary stage of the design, to account for the various technical and economical aspects, their implications and relationships. Results are presented of computations carried out in a specific case, identified by the annual gas production and other characteristics of the well terminal and a cruising route for the ships. The comparison is carried out on the basis of the cost per transported unit of gas and of the percentage of success in the transportation process. The computations show that the choice of the PV material has a key influence on the results in terms of optimal number, dimensions and speed of the ships.

Thermal History Resulting in the Failure of Lightweight Fully-Wrapped Composite Pressure Vessel for Hydrogen in a Fire Experimental Facility

We must improve our understanding of the thermal behaviour of composite gas storage in the event of fire in order to reduce the risk of bursting. In this research, results of pool fire tests were used to improve understanding of the failure mechanisms of epoxy carbon fibre composite pressure vessels with a polymeric liner (type IV vessel) designed for a working pressure of 70 MPa. The failure mode in a pool fire test depends on the storage design and on the initial pressure of the storage. For instance, for a 100 L type IV storage without any safety system, initial pressures of 70 MPa down to 52.5 MPa result in pressure vessels bursting, and initial pressures of 35 MPa down to 17.5 MPa lead pressure vessels to loss of liner tightness. The occurrence of one mechanism or the other is due to the predominance of either heat transfer through the wall, leading to a loss of tightness; or of the degradation of the materials, leading to bursting. Thermogravimetric analyses were carried out on the pressure vessel materials to determine the onset of degradation take during pool fire tests. The temperature measurements allowed for proper characterisation of the conditions leading to the loss of liner tightness. Temperature profiles were used to link the position of the composite degradation front to the loss of tensile strength leading to bursting.

Chemical composition effect on VVER-1000 RPV weld metal thermal aging

Temperature and fast neutron flux simultaneously affect the material of welded joints of reactor pressure vessels under irradiation. Understanding thermal aging effects on the weld metal allows for an explanation of the mechanisms that govern an increase in the ductile-to-brittle transition temperature of the reactor pressure vessel materials under long term irradiation at operation temperature. This paper reports on new results and reassessment of the VVER-1000 weld metal surveillance specimen database performed at the National Research Center “Kurchatov Institute”. The current database of VVER-1000 weld metal thermal aging at 310–320 °C includes 50 transition temperature values with the maximum holding time of 208,896 h. The updated database completed with the information on intergranular fracture shear and phosphorous content in the grain boundaries has allowed us to propose a new mechanism of VVER-1000 weld materials thermal aging at 310–320 °C and develop models of ductile-to-brittle transition temperature shift for VVER-1000 weld metal during a long-term exposure at 310–320 °C.

Methodology of Including Spatial Nonuniformity of Material Properties in the Brittle Fracture Resistance Calculation for WWER-1000 Reactor Pressure Vessels. Part 1. Theoretical Analysis

The paper addresses the main methods of predicting the temperature dependence of fracture toughness. The authors have analyzed a system of margins which have to be introduced in order to adequately and conservatively predict this dependence using the results of testing surveillance coupons. A methodology is put forward for determining a margin for spatial nonuniformity of properties of the reactor pressure vessel material, which is based on simultaneous deterministic and probabilistic calculations of brittle fracture resistance of a reactor pressure vessel.

Comparison of fracture properties in SA508 Gr.3 and Gr.4N high strength low alloy steels for advanced pressure vessel materials

Nuclear power systems are moving to a larger capacity or smaller modular type. In any either case, advanced pressure vessel materials with high strength and toughness are definitely needed for an optimization of the design and construction, as well as the long-term operation. In this paper, two candidate materials, both of which are within the current ASME specifications of SA508 steel forging, are compared from the view point of fracture resistance properties for a nuclear pressure vessel steel. The microstructure and mechanical properties of SA508 Gr.3 Cl.1, Cl.2, and Gr.4N steels were also characterized. The predominant microstructure of SA508 Gr.4N model alloy is tempered martensite, while SA508 Gr.3 Cl.1 and Cl.2 steels show a tempered upper bainitic structure. SA508 Gr. 4N model alloy showed the best strength and transition behavior among the three types of SA508 steel. SA508 Gr.3 Cl.2 steel has good strength and fracture toughness, but there is a decrease in the upper-self energy. The fracture resistance and fatigue crack growth rate of SA508 Gr.3 Cl.2 and Gr.4N steels were comparable to those of SA508 Gr.3 Cl.1 steel. In terms of mechanical properties, SA508 Gr.4N steel is a fascinating material for the pressure vessel application although it still needs verification on the aging behavior such as the irradiation embrittlement resistance.

Effect of a Single Overload on the Fracture Behavior in Safe-End Dissimilar Metal Welded Joints in Nuclear Power Plant

To understand the effect of a single overload on the fracture behavior in welded joints, the stress and strain field at the crack tip in a safe-end dissimilar metal welded joint in nuclear pressure vessel is simulated and analyzed by using the elastic-plastic finite element method in the paper, in which the mechanical heterogeneity in welded joint is emphatically considered. The investigating results indicate that the tensile plastic strain at crack tip increases, but the tensile stress decreases as a single overload increases, and the influence of a single overload on tensile strain is larger than one on tensile stress, which provide a theoretical basis for quantitatively estimating the crack growth rate of environmentally assisted cracking in the welded structural material of pressure vessel and piping in the nuclear power plant.

Effects of fracture toughness curves of ASME Section XI–Appendix G on a reactor pressure vessel under pressure–temperature limit operation

According to the Code Case N-640 issued in 1999, the fracture toughness requirement of reactor pressure vessel materials in ASME Section XI–Appendix G was amended to the KIC curve. In Taiwan, the present pressure–temperature limit operation curves of normal reactor startup (heat-up) and shut-down (cool-down) for the reactor pressure vessel is still calculated per the KIa curve in 1998 or earlier editions. In the paper, the failure risks of a Taiwan domestic reactor pressure vessel under various pressure–temperature limit operations were analyzed. First, the pressure–temperature limit curves of the reactor pressure vessel based on KIa and KIC curves, and various levels of radiation embrittlement, were established. Then, the ORNL's probabilistic fracture mechanics code, FAVOR, and the PNNL's flaw model were employed to assess the failure probabilities of the reactor pressure vessel under such pressure–temperature limit transients. Further, the deterministic analyses of FAVOR code were also conducted. It is found that under the pressure–temperature limit transients based on KIC curves, the reactor pressure vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the pressure–temperature limits can increase the operational margin as well as maintaining the sufficient stability of the analyzed reactor pressure vessel.

Effect of Neutron Irradiation on Brittle Fracture Initiation in VVER-1000 Reactor Pressure Vessel Materials

Abstract This work contains fractographic analysis of base metal and weld metal specimens of a VVER-1000 reactor pressure vessel (RPV) in initial state and after neutron irradiation. In this work, two main types of brittle fracture origins were discovered for VVER-1000 RPV materials in different states—non-metallic inclusions and grain or subgrain boundaries, as well as the correlation between the cleavage initiation distance and critical stress-intensity factor KJc values for each type of fracture origin. In this paper, evolution of relative proportion between these two types of origins under neutron irradiation was also studied.

English

In an autofrettage process a given thick cylinder is subjected to such a pressure which gives a specified depth of elasto-plastic boundary. The outside diameter expansion during autofrettage process is a function of depth of autofrettage. To obtain a specified depth of elastic-plastic interface, the applied autofrettage pressure increases in direct proportion to the proof stress of the pressure vessel material. Although this increases the load bearing capacity of the barrel, resulting in enhanced factor of safety, this increases the maintenance cost of a hydraulic autofrettage plant. To assure quality, product safety and manufacturing economy, an optimal autofrettage pressure is defined. The paper proposes that the minimum autoftrettage pressure is the pressure at which the pressure exterior expansion curve intersects line of constant factor of safety. At higher values of 0.2 per cent proof stress of tube material autofrettaging based on line of constant factor of safety will result in a reduction in fatigue life. The point of intersection of exterior expansion curve with line of constant fatigue life has been defined as the optimal pressure because the specifications of factor of safety and fatigue life are simultaneously achieved. The proposed process design based on above concept has been validated using finite element simulation and empirical post-autofrettage measurements. The verification of the shakedown condition for reverse yielding due to the Bauschinger effect (Huang’s model) and fatigue life has also been satisfied. Defence Science Journal, Vol. 64, No. 4, July 2014, pp. 385-392, DOI:http://dx.doi.org/10.14429/dsj.64.3886

EFFECTS OF TEMPERING AND PWHT ON MICROSTRUCTURES AND MECHANICAL PROPERTIES OF SA508 GR.4N STEEL

Presented in this study are the variations of microstructures and mechanical properties with tempering and Post-Weld Heat Treatment (PWHT) conditions for SA508 Gr.4N steel used as Reactor Pressure Vessel (RPV) material. The blocks of model alloy were austenitized at the conventional temperature of 880 °C, then tempered and post-weld heat treated at four different conditions. The hardness and yield strength decrease with increased tempering and PWHT temperatures, but impact toughness is significantly improved, especially in the specimens tempered at 630 °C. The sample tempered at 630 °C with PWHT at 610 °C shows optimum mechanical properties in hardness, strength, and toughness, excluding only the transition property in the low temperature region. The microstructural observation and quantitative analysis of carbide size distribution show that the variations of mechanical properties are caused by the under-tempering and carbide coarsening which occurred during the heat treatment process. The introduction of PWHT results in the deterioration of the ductile-brittle transition property by an increase of coarse carbides controlling cleavage initiation, especially in the tempered state at 630 °C.

Study of the Evolution of Defects in the Structure of Reactor Pressure Vessel by Rate Theory

In this work, we implemented a chemical rate theory model for the growth of nano-sized point defect clusters (PDCs) and copper- rich precipitates (CRPs) which can change the mechanical properties in the reactor pressure vessel material of a nuclear power plant. For the calculation of irradiation defect evolution, a number of time-dependent differential equations were established and numerically integrated. The concentration of mono-size vacancies and interstitials was saturated at an early stage of irradiation, and it was found that the vacancy concentration was higher than the interstitial concentration. The high concentration of vacancies induced a growth of the CRPs at the later stage. The concentration of PDCs and the size of CRPs were used to estimate the mechanical changes, and the calculation results were compared with the measured changes in yield strength and Charpy V-notch transition temperature shift obtained from the surveillance test data of Korean light water reactors (LWRs). It was observed that the estimated values were in fair agreement with the experimental results in spite of the uncertainty regarding the material property parameters and modeling method.

Failure of a pressure vessel for rail transport of fluid carbon dioxide

In 1976, a pressure vessel of a tank wagon burst during gravity shunting. The pressure vessel was filled with cold fluid carbon dioxide and failed by cleavage fracture explosively. Big parts of the vessel were flung up to 300m, a building was destroyed, and rails as well as the superstructure of the gravity shunting yard were damaged.The initiation sites of the fracture and the orientations of crack propagation were determined by means of the patterns on the fracture surfaces of the broken parts. The initiation sites were located in the heat-affected zone (HAZ) of welds, close to transitions of cross-sections and the starting and end points, respectively, of the welds. The primary initiation of crack propagation was almost certainly on the left side, close to the weld of the fixed bearing connecting the vessel with the carriage.The results of metallographic investigations indicated the development of cold weld cracks during welding. Mechanical behavior assessments on specimens extracted from the failed vessel indicated sufficient strength and ductility in tensile loading. The values of the notched-bar impact energy or toughness were within the limits according to industry rules and standards, except for one metal sheet of the vessel that was welded to the left fixed bearing.During shunting, two tank wagons were inappropriately braked, and began to roll too quickly. One rail brake used for deceleration of the wagons caused impulsive noises and vibrations, resulting in high impulsive loading of the first wagon. The cold weld cracks in the heat-affected zones near the fixed bearing of the pressure vessel were initiation points for unstable crack growth. As the result of very low notched-bar impact toughness, crack propagation was not arrested and unstable cleavage fracture occurred.Due to a second similar accident, happened one year before, a program was started to assess characteristic values of pressure vessel materials and the influence of weld parameters by the manufacturer, the inspecting authority, and BAM. The scientific findings led to modifications of industry rules and standards with regard to using ISO-V specimens and to higher required values for notched-bar impact energy for fine grain structural steels.

Neutron fluence analyses around the reactor pressure vessel of BWR using MCNP with a heterogeneous and homogeneous mixed core model

For reactor pressure vessel (RPV) material surveillance program, it is necessary to obtain fast neutron fluence. In this work, Monte Carlo transport code MCNP is applied to analyze it for a Boiling Water Reactor (BWR) with a heterogeneous and homogeneous mixed core model (HHMCM) and its applicability is examined. The analyses of using MCNP with HHMCM are performed to obtain the neutron flux and the reaction rate of the dosimeter wires at the inner surface of the RPV of an existing 800 MWe BWR plant in Japan. As a result, the neutron flux and the reaction rates can be estimated with an uncertainty of 8% at most. In addition, HHMCM can reduce a calculation time to 1/9 compared with a case of all bundles treated as heterogeneous.