Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550 °C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550 °C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550 °C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L.     

Stress corrosion cracking of Alloy 600 in high temperature water

Slow strain rate stress corrosion tests have been performed on specimens cut from four separate heats of Alloy 600 steam generator tubing. The material was tested in the mill-annealed and thermally-stabilised conditions and after various low temperature ageing treatments. Only limited cracking was observed, even for tests at 340°C, but the initiation of intergranular cracking was easier on the inner than on the outer surfaces on the tubing. Polarization data has been obtained in high-temperature water and in saturated boric acid and saturated lithium hydroxide at the atmospheric boiling points, and slow strain tests were performed at controlled potentials in these environments. Again, only very short cracks formed during the slow strain rate tests which were performed at a strain rate of about 10⁻⁶ s⁻¹. The data is discussed in terms of the probable crack tip strain rates that would exist in these tests and at other strain rates. It is argued that if cracking occurs, the main role of very slow strain rate tests is to provide time for initiation and crack growth, so that cyclic loading or intermittent loading long term tests are likely to be more successful in sustaining crack growth in this alloy.     

Stress corrosion cracking of low-alloy, reactor-pressure-vessel steels in oxygenated, high-temperature water

The stress corrosion cracking (SCC) behaviour of low-alloy, reactor-pressure-vessel (RPV) steels in oxygenated, high-temperature water and its relevance to boiling water reactor (BWR) power operation, in particular its possible effect on both RPV structural integrity and safety, has been a subject of controversial discussions for many years. This paper presents the results of an experimental study on crack growth through SCC in three, nuclear-grade, steels (SA 533 B Cl.1, SA 508 Cl.2, 20 MnMoNi 5 5) under simulated, BWR water-chemistry conditions. Modern, high-temperature water loops, on-line crack-growth monitoring and fractographic analysis in the scanning electron microscope were used to quantify the cracking response of pre-cracked, fracture-mechanics specimens under a variety of mechanical and environmental conditions. Corrosion-assisted crack advance could be only initiated by active loading within the environment. If SCC crack advance at constant load was observed, initiation of crack growth had always occurred while increasing the load to the intended value for subsequent, static-load testing. During the constant load period the rate of SCC crack advance rapidly decayed and crack arrest occurred within a period of <100 h (for tests with K_I60 MPa m^1/2). Supplementary experiments with slowly increasing loading revealed that the initiation of crack growth, and the extent of further crack advance, are crucially dependent upon maintaining both a positive crack-tip strain rate and a high sulphur-anion activity in the crack-tip environment. It is concluded that there is no sustainable susceptibility to SCC crack growth under purely static loading, as long as small-scale-yielding conditions prevail at the crack-tip and the water chemistry is maintained within current BWR/NWC operational practice (EPRI water chemistry guidelines). However, sustained, fast SCC (with respect to operational time scales) cannot be excluded for faulted water-chemistry conditions (>EPRI action level 3) and/or for highly stressed specimens either loaded near to K_IJ or with a high degree of plasticity in the remaining ligament.     

Stress corrosion cracking in the heat-affected zone of A508 steel welds under high-temperature water

The influence of heat input on stress corrosion cracking (SCC) in the heat-affected zone (HAZ) of A508 steel welds was investigated. Constant extension rate tensile tests were conducted on notched round-bar specimens in simulated reactor coolant conditions to assess SCC performance. In multi-pass welds, the use of a low heat input resulted in a better SCC resistance than that of a high heat input due to the existence of a more refined microstructure.     

Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions

The stress corrosion cracking (SCC) behaviour of different reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated boiling water reactor (BWR) normal water (NWC) and hydrogen water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (<0.6 mm/year) under most BWR/NWC and material conditions. Critical water chemistry, loading and material conditions, which can result in sustained and fast SCC well above the ‘BWRVIP-60 SCC disposition lines’ were identified, but many of them generally appeared atypical for current optimized BWR power operation practice or modern RPVs. Application of HWC always resulted in a significant reduction of SCC crack growth rates by more than one order of magnitude under these critical system conditions and growth rates dropped well below the ‘BWRVIP-60 SCC disposition lines’.     

Stress corrosion cracking of Zircaloys in unirradiated and irradiated CsI

Unirradiated split-ring specimens of Zircaloy fuel cladding, coated with CsI, cracked when stressed at elevated temperatures. The specimens have been reexamined fractographically and metallographically in order to confirm that the cause of cracking was stress corrosion (SCC) and not delayed hydride cracking (DHC). Further specimens have been cracked at 350°C by a solution of CsI in a fused mixture of nitrates of rubidium, cesium, strontium and barium, by a similar mechanism. CsI dissolved in a fused molybdate melt was not stable at 400°C, and rapidly evolved iodine, leaving a melt that was incapable of causing SCC. Irradiation of stressed split-ring specimens of Zircaloy fuel cladding in a γ-irradiator of 10⁶ R/h and in the U-5 loop in the NRU reactor at an estimated 10⁹ R/h caused SCC when the specimens were packed in dry CsI powder. Care had to be taken to dry the CsI, otherwise cracking occurred by a DHC mechanism from hydrogen absorbed from residual moisture in the CsI. Fractography showed that the crack surfaces obtained with dry CsI were typical of iodine-induced SCC rather than cesium-induced metal vapour embrittlement. Thus, if a transport process is provided for the iodide to obtain access to the zirconium surface, CsI is capable of causing SCC of Zircaloy. This transport process might be ionic diffusion in a fission product oxide melt in the fuel-clad gap, however, radiolysis of CsI to form a volatile iodine species in a radiation field is the more probable explanation of PCI failures.     

Corrosion behavior of porous chromium carbide in supercritical water

The corrosion behavior of highly porous chromium carbide (Cr₃C₂) prepared by a reactive sintering process was characterized at temperatures ranging from 375 °C to 625 °C in a supercritical water environment with a pressure of 25–30 MPa. The test results show that porous chromium carbide is stable in SCW environments at temperatures under 425 °C, above which disintegration occurred. The porous carbide was also tested under hydrothermal conditions of pressures between 12 MPa and 50 MPa at constant temperatures of 400 °C and 415 °C, respectively. The pressure showed little effect on the stability of chromium carbide in the tests at those temperatures. The mechanism of disintegration of chromium carbide in SCW environments is discussed.     

Corrosion behavior of Alloy 625 in supercritical water environments

For a better understanding on the corrosion behavior of Alloy 625, samples fabricated from this alloy were exposed to supercritical water (SCW) environments with 8.3 ppm dissolved oxygen at 400 and 600 °C and 24.8 MPa (3600 psi) for various periods of time up to 1000 h. Pits were found on the surfaces of the samples after the corrosion tests, and the formation of these pits could be attributed to metal carbide inclusions in the as-received Alloy 625. Mass changes (w) in the samples as a function of test duration (t) could be fitted by an equation of w^2.21 = 1.4 × 10⁻⁵ t, indicating that the mass change approximately followed the parabolic law in the specified SCW environment. In addition, oxides with a double-layer structure were observed on the samples. The outer layer of the oxides consisted mainly of sub-micron spinels of Ni(Cr,Fe)₂O₄, and the compact inner layer was mixed Ni(Cr,Fe)₂O₄ and Cr₂O₃ with a grain size of tens of nanometer.     

CFD analysis of thermal-hydraulic behavior of supercritical water in sub-channels

Investigations on the thermal-hydraulic behavior in the supercritical water-cooled reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical fluids. In this paper, computational fluid dynamics (CFD) analysis is carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical SCWR fuel assembly using commercial CFD code CFX-5.6. Three types of sub-channels, e.g. regular sub-channel, wall sub-channel and corner sub-channel, are analyzed. Effects of various parameters, such as boundary conditions and pitch-to-diameter ratios, on the mixing phenomenon in sub-channels and heat transfer are investigated. The turbulent mixing in tight lattice (P/D = 1.1) is lower than that in wide lattice (P/D > 1.1), whereas, the effect of pitch-to-diameter ratio on the turbulent mixing is slight at P/D > 1.1. The amplitude of turbulent mixing in wall sub-channel is slightly higher than that in regular sub-channel and is close to that in corner sub-channel. The mixing coefficient in the sub-channel at P/D ≥ 1.2 is in the range from 0.022 to 0.028. The results also show unusual behavior of turbulent mixing in the vicinity of the pseudo-critical point, and further investigation is needed. The mass mixing due to cross flow in wall sub-channel is much stronger than that in regular sub-channel at a same pitch-to-diameter ratio. The mass mixing in wall and regular sub-channels, especially at small pitch-to-diameter ratio, brings an unfavorable feedback to the heat transfer and strengthens the non-uniformity of the circumferential distribution of heat transfer. The strong mass mixing in corner sub-channel should be paid attention.     

Stress corrosion cracking of stainless-steel canister for concrete cask storage of spent fuel

Resistance to external stress corrosion cracking (ESCC) and crevice corrosion were examined for various candidate canister materials in the spent fuel dry storage condition using concrete casks. A constant load ESCC test was conducted on the candidate materials in air after deposition of simulated sea salt particles on the specimen gage section. Highly corrosion resistant stainless steels (SS), S31260 and S31254, did not fail for more than 46 000 h at 353 K with relative humidity of 35%, although the normal stainless steel, S30403 SS failed within 500 h by ESCC. Crevice corrosion potentials of S31260 and S31254 SS became larger than 0.9 V (SCE) in synthetic sea water at temperatures below 298 K, while those of S30403 and S31603 SS were less than 0 V (SCE) at the same temperature range. No rust was found on S31260 and S31254 SS specimens at temperatures below 298 K in the atmospheric corrosion test, which is consistent with the temperature dependency of crevice corrosion potential. From the test result, the critical temperature of atmospheric corrosion was estimated to be 293 K for both S31260 and S31254 SS. Utilizing the ESCC test result and the critical temperature, together with the weather station data and the estimated canister wall temperature, the integrity of canister was assessed from the view point of ESCC.     

Quantitative examination of stress corrosion cracking of alloy 600 in high temperature water

Stress corrosion cracking (SCC) examination of Inconel 600 steam generator tubing has continued at Brookhaven National Laboratory, using U-bends, constant load and slow extension rate tests, leading to Arrhenius plots of failure times versus inverse temperature for crack initiation and propagation. Effect of applied load can be expressed as log-log curves for failure times versus stress. Variations in environment and cold work are included in all the experiments. Microstructure and composition of oxide films on Inconel 600 surfaces were examined after exposure to pure water at 365°C, and stripping with the bromine-methanol method. Results are consistent with a mechanism of transient creep, film rupture and a mass-transport-limited anodic process.     

Simulation of stress corrosion cracking behavior in a tube-shaped specimen of nickel-based alloy 600

Stress corrosion cracking (SCC) simulation code has been developed for the evaluation of SCC behavior in specimens in the shape of field components. The code utilizes numerical calculation of stress/strain states at a crack tip using finite element methods and a formula describing the crack tip reaction kinetics containing unknown environmental parameters. The applicability of this simulation code was investigated by applying the code to the evaluation of SCC behavior in a mock-up of a bottom mounted instrumentation tube for a pressurized water reactor subjected to complex stress/strain states. The results indicate that crack growth rate in a component suffering from certain environments can be estimated using the developed SCC simulation code with pre-determined unknown parameters, using the experimental crack growth rate data measured on other specimens in the same environment.     

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 dK_I/dt, 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.     

Effect of microstructure on the corrosion of CVD-SiC exposed to supercritical water

Silicon carbide (SiC) is an important engineering material being studied for potential use in multiple nuclear energy systems including high-temperature gas-cooled reactors and water-cooled reactors. The corrosion behavior of SiC exposed to supercritical water (SCW) is critical for examining its applications in nuclear reactors. Although the hydrothermal corrosion of SiC has been the subject of many investigations, the study on the microstructural effects on the corrosion is limited. This paper presents the effect of residual strain, grain size, grain boundary types, and surface orientations on the corrosion of chemical vapor deposited (CVD) β-SiC exposed to SCW at 500 °C and 25 MPa. Weight loss occurred on all the samples due to localized corrosion. Residual strains associated with small grains showed the most significant effect on the corrosion compared to the other factors.     

Advanced TEM characterization of stress corrosion cracking of Alloy 600 in pressurized water reactor primary water environment

Advanced transmission electron microscopy techniques were carried out in order to investigate stress corrosion cracking in Alloy 600 U-bend samples exposed in simulated PWR primary water at 330 °C. Using high-resolution imaging and fine-probe chemical analysis methods, ultrafine size oxides present inside cracks and intergranular attacks were nanoscale characterized. Results revealed predominance of Cr₂O₃ oxide and Ni-rich metal zones at the majority of encountered crack tip areas and at leading edge of intergranular attacks. However, NiO-structure oxide was predominant far from crack tip zones and within cracks propagating along twin boundaries and inside grains. These observations permit to suggest a mechanism for intergranular stress corrosion cracking of Alloy 600 in PWR primary water. Indeed, the results suggest that stress corrosion cracking is depending on chromium oxide growth in the grain boundary. Oxide growth seems to be dependent on oxygen diffusion in porous oxide and chromium diffusion in strained alloy and in grain boundary beyond crack tip. Strain could promote transport kinetic and oxide formation by increasing defaults rate like dislocations.     

不同通道内超临界水流动换热特性

以具有不同当量直径的矩形、圆形、三角形、环形等管道为研究对象,利用ANSYS CFX分别计算在定质量流量和定流体速度条件下,超临界水在不同通道内的流动换热特性。发现在定质量流量条件下,圆形通道换热特性最优,且小当量直径促进流体换热;在定入口流速条件下,环形通道换热特性最优,小当量直径管道促进亚临界流体换热,大当量直径管道促进超临界流体换热。     

Stress corrosion cracking countermeasure observed on Ni-based alloy welds of BWR core support structure

The effect of hydrostatic test on the residual stress re-distribution was simulated by experiment to confirm the residual stress behavior of the cone-shaped shroud support to reactor pressure vessel (RPV) weld, where a number of cracks due to stress corrosion cracking (SCC) were observed on the inner side only. Test specimen with tensile residual stress was loaded and unloaded with axial plus bending load, which simulates the hydrostatic test load, and the strain change was measured during the test to observe the residual stress behavior. The results verify that the residual stresses of the shroud support to the RPV weld were reduced and the stresses on inner and outer sides were reversed by the hydrostatic test. As the SCC countermeasure, the shot peening (SP) technology was applied. Residual stress reduction by SP on the complicated configuration, and improvement of SCC resistance and endurance of the compressive residual stress were experimentally confirmed. Then, SP treatment procedures on the actual structure were confirmed and a field application technique was established.     

Provisions to prevent strain-induced corrosion cracking in ferritic piping systems of boiling water reactors

Based on failure analysies and evaluations of laboratory test results, system areas are localized which are to be regarded as susceptible to strain-induced corrosion cracking (SICC). Possible systems engineering, operational, design and manufacturing provisions for preventing SICC are described. Necessary measures have already been taken by BWR operators, acting in part together with authorised experts and licensing authorities, in the case of plants which are in operation.     

Oxidation behavior of Incoloy 800 under simulated supercritical water conditions

For a correct design of supercritical water-cooled reactor (SCWR) components, data regarding the behavior of candidate materials in supercritical water are necessary. Corrosion has been identified as a critical problem because the high temperature and the oxidative nature of supercritical water may accelerate the corrosion kinetics. The goal of this paper is to investigate the oxidation behavior of Incoloy 800 exposed in autoclaves under supercritical water conditions for up to 1440 h. The exposure conditions (thermal deaerated water, temperatures of 723, 773, 823 and 873 K and a pressure of 25 MPa) have been selected as relevant for a supercritical power plant concept. To investigate the structural changes of the oxide films, X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive X-ray spectrometry (EDX) and electrochemical impedance spectroscopy (EIS) analyses were used. Results show changes in the oxides chemical composition, microstructure and thickness versus testing conditions (pressure, temperature and time). The oxide films are composed of two layers: an outer layer enriched in Fe oxide and an inner layer enriched in Cr and Ni oxides corresponding to small cavities supposedly due to internal oxidation.