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.     

The feasibility of small size specimens for testing of environmentally assisted cracking of irradiated materials and of materials under irradiation in reactor core

Environmentally assisted cracking (EAC) or, in other words, stress corrosion cracking (SCC) of in-core materials has become an increasingly important reason for the downtime and maintenance costs of nuclear power plants (NPPs). Use of small size specimens for stress corrosion testing of irradiated materials is necessary because handling of high dose rate materials is difficult and the availability of these materials is limited. A drawback of using small size specimens is that they do not in some cases fulfil the requirements of the relevant testing standards and sometimes their limited load-bearing capacity prevents corrosion fatigue tests and tests with static loading at reasonable K_I values. The test results show that the ductile fracture resistance curves of a Cu–Zr–Cr alloy are, to some extent, independent of the specimen geometry and size. However, the curves of small specimens deviate from the curves of larger specimens at high J values (large plastic zone relative to the remaining ligament) or when the crack growth exceeds about 30% of the remaining ligament. The size dependency of the tested Cu–Zr–Cr alloy seems to be a consequence of decreasing stress triaxiality as the size of the specimen is decreased. The results of the SCC tests of sensitized SIS 2333 stainless steel (equal to AISI 304) specimens in simulated boiling water reactor (BWR) water show that the plastic deformation of the remaining ligament of the specimen has no significant effect on the environmentally assisted crack growth rate. This indicates that stress corrosion testing is not limited by the specimen size. The size dependency in SCC tests should be further studied by conducting tests using various specimen sizes.     

Effect of environment, mechanical conditions, and materials characteristics on AE behavior during corrosion fatigue processes of an austenitic stainless steel

Acoustic emission (AE) behavior during corrosion fatigue (CF) processes has been studied in a commercial grade 304 austenitic stainless steel with special attention to the effect of environment, mechanical conditions, and materials characteristics. Precracked compact tension specimens were tested under cyclic tension-tension load, polarized potentiostatically in 1N H₂SO₄ + 0.5M HaCl or 1N H₂SO₄ solutions at room temperature. Valuable AE signals discriminated from environmental and mechanical noises by source location were monitored at the loading phase near the maximum load. The experimental results showed that the AE activity observed in corrosive environments increased significantly with the acceleration of crack growth rates, compared with that observed in air (inert environment). Detailed SEM observations found cleavage-like cracking, intergranular-like cracking, separations, etc. on the crack surfaces. The AE sources which provided the high AE activity during CF crack growth were shown to be such microcracking processes as cleavage-like cracking, intergranular-like cracking, and separations caused by the cooperation of environments (intergranular corrosion or hydrogen), mechanical conditions (triaxial stress), and materials characteristics (non-metallic inclusions or carbides precipitated along grain boundaries by sensitization). The effect of these three factors on the AE behavior and cracking mechanisms is discussed in detail.     

Positron annihilation lifetime study of JRQ and JPA irradiated and annealed reactor pressure vessel steels

The model reactor pressure vessel steels known as JRQ and JPA were manufactured in Japan for the IAEA neutron embrittlement research studies. These model alloys belong to the commercially used steel A533B-1 type and show relatively large changes in mechanical properties after relevant neutron irradiation. The neutron irradiation was performed by different neutron fluxes as well as different neutron fluences (up to about 150 × 10¹⁸ cm⁻² (E > 0.5 MeV)). For a better understanding of the neutron embrittlement, the Positron Annihilation Lifetime Spectroscopy (PALS) technique was applied in 2014. PALS measurement of irradiated specimens was performed using three detectors set-up due to induced ⁶⁰Co radioactivity of the studied specimens. We confirmed that the JPA steel, considered to be high-copper steel, is much more sensitive to defect creation due to neutron irradiation than the low-copper JRQ steel.     

The effects of environment,orientation and strength level on the stress corrosion behavior of U-0.75 wt % Ti

The effects of the environment, specimen orientation and the alloy strength level on the susceptibility of U-0.75 wt % Ti to stress corrosion crack propagation have been determined. The data show that water (H₂) is the species responsible for cracking and the Cl^? concentration has little effect on the cracking behavior. The orientation of extruded specimens had little effect on the cracking behavior. The data also show that as the strength level of the alloy increases due to varying the aging treatment the threshold stress intensity for stress corrosion crack propagation decreases linearly.     

Environmentally assisted crack growth in a low alloy boiler steel in high temperature water containing oxygen

The crack growth behaviour of the higher strength steel 17 MnMoV 6 4 in deionized high temperature water containing oxygen was investigated with respect to static loading. The tests were performed in an autoclave testing facility at an oxygen content of 8 ppm, a pressure of 70 bar and a temperature of 240°C under quasi-stagnant flow conditions. The stress intensities at the beginning of the tests were 17, 20, 27, 34, 40 and 58 MPa√m. In contrast to the higher loaded specimens no evidence of environmentally assisted cracking was found at stress intensities of 17 and 20 MPa√m.The maximum crack growth rate in the range where only environmetally assisted cracking occured amounted to about 4 x 10^-5 mm/s. The critical stress intensity K_IJ for the onset of stable ductile crack growth in air is 100 MPa√m. On the basis of fractographic studies the crack development found can be ascribed to the “Strain Induced Corrosion Cracking (SICC)” mechanism. This mechanism, used as a working hypothesis, gives a satisfactory explanation for the occurence of stress corrosion cracking of unalloyed and low alloyed steels in high temperature water. SICC is particularly characterized by aggravated corrosive attack occuring as soon as the magnetite/haematite protective layer has been locally disturbed. The stress concentration then just becomes so great that in the region of the resulting crack tips, yield/creep deformation within the critical range of strain rates occurs.     

In-Core SCC Growth Behavior of Type 304 Stainless Steel in BWR Simulated High-Temperature Water at JMTR

Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact tension-type specimens of type 304 stainless steel that had been pre-irradiated up to a neutron fluence level around 1 × 10²⁵ n/m² under a pure water simulated boiling water reactor (BWR) coolant condition at the Japan Materials Testing Reactor (JMTR). In order to investigate the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate, we performed ex-core IASCC tests on irradiated specimens at several dissolved oxygen contents under the same electrochemical potential condition. In this paper, results of the in-core SCC growth tests are discussed and compared with the results obtained by ex-core tests from a viewpoint of the synergistic effects on IASCC. From results of in-core and ex-core tests using pre-irradiated specimens, the effect of synergy of neutron/gamma radiation and stress/water environment on SCC growth rate was considered to be small, because the in-core data under the same ECP condition were similar to the ex-core data under the DO = 32 ppm condition.     

European experiences of the proposed ASTM test method for crack arrest toughness of ferritic materials

The proposed ASTM test method for measuring the crack arrest toughness of ferritic materials using wedge-loaded, side-grooved, compact specimens was applied to three steels: A514 bridge steel tested at −30°C (CV30–50°C), A588 bridge steel tested at −30°C (CV30–65°C), and A533B pressure vessel steel tested at +10°C (CV30-12°C) and +24°C (CV30+2°C). Five sets of results from different laboratories are discussed here; in four cases FOX DUR 500 electrodes were used for notch preparation, in the remaining case HARDEX-N electrodes were used. In all cases, notches were prepared by spark erosion, although root radii varied from 0.1–1.5 mm. Although fast fractures were successfully initiated, arrest did not occur in a significant number of cases.The results showed no obvious dependence of crack arrest toughness, K_a, (determined by a static analysis) on crack initiation toughness, K₀. It was found that K_a decreases markedly with increasing crack jump distance, Δα/W. A limited amount of further work on smaller specimens of the A533B steel showed that lower K_a values tended to be recorded.It is concluded that a number of points relating to the proposed test method and notch preparation are worthy of further consideration. It is pointed out that the proposed validity criteria may screen out lower bound data. Nevertheless, for present practical purposes, K_a values may be regarded as useful in providing an estimate of arrest toughness — although not necessarily a conservative estimate.     

Corrosion fatigue behavior of low alloy steels under simulated BWR coolant conditions

The corrosion fatigue crack growth behavior of A533 and A508 low alloy steels under simulated boiling water reactor (BWR) coolant conditions was studied. Corrosion fatigue crack growth rates of A533B3 and A508 cl. 3 steels were significantly affected by the steel sulfur content, loading frequency and dissolved oxygen content of water environments. The data points outside the bound of Eason’s model could be attributed to the low frequency, higher steel sulfur content and high dissolved oxygen in water environments. The sulfur dissolved in the water environment from the higher-sulfur steels was sufficiently concentrated to acidify the crack tip chemistry even in the hydrogen water chemistry (HWC). Therefore, nitrogenated or HWC water showed little or no beneficiary effect on the high-sulfur steels. For the steel specimens of the same sulfur level, their corrosion fatigue crack growth rates were comparable in different orientations, which could be related to the exposure of fresh sulfides to the water environment. The percentages of sulfides per unit area, by quantitative metallography, were comparable for the steel specimens of both orientations. When the steel sulfur content was decreased to a critical sulfur content 0.005 wt.%, the crack growth rates decreased remarkably.     

Modeling of fatigue crack growth rate for ferritic steels in light water reactor environments

This paper presents the methodology and results of an effort to analyze and model a large set of fatigue crack propagation data on A508 Class 2 and Class 3 and A533B pressure vessel steel in light water reactor (LWR) environments. The data were from a variety of laboratories worldwide, in most cases contributed to the EPRI Database on Environmentally Assisted Cracking (EDEAC). The data were analyzed in a consistent manner using the computer code FATDAC, which minimizes the scatter arising from numerical differentiation during fatigue data reduction. The models were developed in the time domain, then converted to the more conventional da/dN versus ΔK form. Two modes of corrosion fatigue crack growth behavior were identified and modeled, one about a factor of two faster than the rates in air and independent of loading rate and the other up to two orders of magnitude faster and strongly dependent on loading rate. Variables such as material sulfur content, sulfide inclusion morphology, water chemistry, R-ratio, load rise time, stress intensity range, temperature, electrochemical potential, and flow velocity affect both the probability of observing the highly enhanced crack growth rates and the rates themselves. Representative crack propagation models are developed and presented in this paper together with supporting data.     

Effects of Hydrogen Peroxide on Intergranular Stress Corrosion Cracking of Stainless Steel in High Temperature Water, (III)

The stress corrosion cracking (SCC) of structural materials used in boiling water reactors has been studied at relatively low hydrogen peroxide (H₂O₂) concentrations, around lOppb, which was assumed to be representative of the corrosion environment formed in hydrogen water chemistry (HWC). The 1/4T compact tension specimen was used for measurement of crack growth rates (CGRs) of sensitized type 304 stainless steel in high temperature and high purity water. Crack length was monitored by a reversing direct current potential drop method. Since H2O2 is easily decomposed thermally, a polytetrafluoroethylene-lined autoclave was used to minimize its decomposition on the autoclave surface. The CGR in the H₂O₂ environment differed from that in the O₂ environment even though the electrochemical corrosion potential (ECP) for both conditions was the same. The data implied that the ECP could not be used as a common environmental deterministic parameter for SCC behavior at higher potentials for different oxidant conditions. The corrosion current density was found to play an important role as an environmental index for SCC, which was given as just the current density at the ECP at a specific oxidant concentration. The CGRs were found to be written as CGR = (3.8±0.6)xl0^-3 _icor +(l-5±1.6) x 10^-8mm/s using the calculated corrosion current density i_corbelow 10^-4 A-cm^-2.     

Fracture initiation and propagation in a cylindrical shell containing an initial surface flaw

The paper deals with the problem of fracture initiation, propagation, and arrest in a pressurized cylindrical vessel which contains an initial surface flaw. It is assumed that the flaw has the most unfavorable geometry and orientation, namely, it is a relatively long part-through axial crack.First we consider the problem of a crack which is sufficiently ‘shallow’ so that the plastic deformations are confined to the neighborhood of the crack border and part of the net section near the inner wall is still elastic. The plasticity-corrected stress intensity factor obtained from this analysis is the controlling load factor in failure considerations related to fatigue crack propagation, stress corrosion cracking, and static fracture (with the use of fracture toughness, COD, or a K_R curvetype failure criterion).The problem of relatively deep crack with fully-yielded net ligament is then considered. Plastic deformations are also assumed to spread around the crack ends through the entire wall thickness. A perfectly plastic strip model (with an eight order shell theory) is used to calculate the plastic zone size and the crack opening displacement along the crack border. Previous studies indicate that for the analysis of the type of stable and subsequent unstable crack propagation problems under consideration, the crack opening displacement δ is a more suitable load factor than the stress intensity factor K, or the crack extension force G. Thus, in this paper a ‘crack opening stretch’ type material characterization will be used.After the rupture of the net ligament under the crack, the axial crack propagation is accompanied by the depressurization of the vessel caused by leakage. From this point on the fracture problem is coupled with the related fluid mechanics or gas dynamics problem where the primary unknowns are the pressure and the crack length as functions of time. In the present study it is assumed that the volume of the vessel is finite and the crack propagation is quasi-static (this assumption, which is necessary to keep the problem within manageable proportions, is justified by the relatively low crack velocities, i.e. v_c < 0.25 c₂, c₂ being the shear wave velocity).     

New observations on the crack growth rate of low alloy nuclear grade ferritic steels under constant active load in oxygenated high-temperature water

Within the scope of reactor safety research attempts have been made over several decades to determine corrosion-assisted crack growth rates. National and international investigations have been performed on both an experimental and an analytical basis. A compilation of internationally available experimental data for ferritic steels exhibits a scatter of crack growth rates of up to 5 decades. This was one of the reasons for commencing further experimental investigations focused on the evaluation of corrosion-assisted crack growth rates. These experimental studies were performed under constant, active, external load on 2T-CT specimens of the materials 20 MnMoNi 5 5 with 0.009 and 0.020% S (similar to A508 Cl.3), 22 NiMoCr 3 7 with 0.006% S (similar to A508 Cl.2) and 17 MnMoV 6 4 with 0.017% S. The tests were carried out in deionized oxygenated high-temperature water (240°C; 0.4 and 8.0 ppm O₂). For K_I values up to 60 MPa m^1/2, the experimental results showed no significant dependence between corrosion-assisted crack growth rates and the stress intensity factor, the oxygen content of the medium or the sulphur content of the steel. Here it is important to note, that in this K_I region the high crack growth rates after the onset of cracking due to loading are decreasing and finally come to a standstill after a short period of time as compared with operational times of plants. Consequently, the determination of crack growth velocities as corrosion-assisted crack advance divided by the test duration, so far practised worldwide, results in wrong crack growth rate values in the above-mentioned range of loading up to 60 MPa m^1/2. Based on a test duration of 1000 h, the average crack growth rates are below 10⁻⁸ mm s⁻¹ for K_I ≤ 60 MPa m^1/2. When applied to a single start-up and service period of one year, this would formally lead to an average crack growth rate of 2·10⁻⁹ mm s⁻¹ (equivalent to 0.06 mm per year). At K_I values between 60 and 75 MPa m^1/2 the average corrosion-assisted crack growth rates increase significantly. It can be observed experimentally that the crack propagates during the whole period of the test. Consequently the calculation of crack growth velocities as corrosion-assisted crack advance divided by the test duration as mentioned earlier can be applied as a first estimate. Finally, for K_I values ≥ 75 MPa m^1/2 high crack growth rates up to 10⁻⁴ mm s⁻¹ can be observed. In this region the average crack growth rates are also in quite good agreement with a theoretically based crack growth model.     

Crack propagation characteristics in neutron irradiated type 316 stainless steel microspecimens

In order to better relate the macroscopic mechanical behavior of irradiated alloys to their associated microstructural condition, unirradiated and neutron irradiated microspecimens were tensile tested at 25–600°C in a quantitative load elongation stage while under continuous observation in a high voltage electron microscope (HVEM). The microtensile specimens, 40 μ m thick, of type 316 stainless steel were irradiated at ambient temperature to a fluence of 1 × 10²² n/m² with 14 MeV neutrons in the Lawrence Livermore Rotating Target Neutron Source II (RTNS) facility.Crack angles, directions and length plotted against total specimen elongation were used to describe the manner in which a crack progressed through each specimen. Rapid crack propagation is accompanied by rapidly changing crack angles and direction and conversely slow propagation corresponds to slowly changing variables. A graph of cumulative crack length plotted against total elongation exhibits a slope which increases as specimen ductility decreases. This graph reflects changes due to the effect of neutron irradiation.     

工质压力及蠕变对冷变形310S不锈钢在超临界水环境下的应力腐蚀开裂行为影响研究

310S不锈钢是一种性能较好的超临界水冷堆候选包壳材料,为丰富310S不锈钢在在超临界水环境下的应力腐蚀性能研究,特别是裂纹扩展速率方面的数据。本研究使用在线监测裂纹扩展的方法,测量了不同冷变形的310S不锈钢在多种工况下的裂纹扩展速率,分析了工质压力、高温蠕变等因素对310S开裂行为的作用。结果显示:超临界水或高温蒸汽的压力变化对310S不锈钢在500℃下的开裂行为的影响较为有限,冷变形作用促进材料的裂纹扩展,材料的高温蠕变行为在超临界水中对应力腐蚀开裂过程中具有较为重要的加速作用,特别是对于高冷变形和高载荷条件下的材料。本研究丰富了超临界水环境下310S的应力腐蚀裂纹扩展速率的数据,证明了提高材料的抗蠕变性能是优化包壳材料服役性能的重要手段之一,包壳设计制造的过程中应当避免较大幅度的冷变形。      

Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

Three-dimensional interpretation of cleavage fracture tests of cladded specimens with local approach to cleavage fracture

Electricité de France has conducted during these last years an experimental and numerical research programme in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels integrity assessment, regarding the risk of brittle fracture. Two cladded specimens made of ferritic steel A508 Cl3 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature to obtain cleavage failure. The crack instability was obtained in base metal by cleavage fracture, without crack arrest. The tests have been interpreted by local approach to cleavage fracture (Beremin model) using three-dimensional finite element computations. After the elastic–plastic computation of stress intensity factor K_J along the crack front, the probability of cleavage failure of each specimen is evaluated using m, σ_u Beremin model parameters identified on the same material. The failure of two specimens is conservatively predicted by both analyses. The elastic–plastic stress intensity factor K_J in base metal is always greater than base metal fracture toughness K_1c. The calculated probabilities of cleavage failure are in agreement with experimental results. The sensitivity of Beremin model to numerical aspects is finally exposed.     

Crack initiation and growth under creep and fatigue loading of an austenitic stainless steel

Both the initiation and the propagation of macroscopic cracks have been studied in a creep ductile 316L type stainless steel at 575–650°C using various fracture mechanics specimens and a wide range of test conditions including steady load at constant or varying temperatures, varying loads at constant temperature. It is shown that, even for isothermal tests, the C^* parameter is unable to provide unique correlations for all the stages of both creep crack initiation and growth. A unique correlation nevertheless exists between C^* and the time to initiation, T_i. Large differences – either conservative or not – from a simplified linear damage cumulation rule are found when the tests are performed at two successive temperatures or two loads. Very detrimental effects of creep-fatigue loadings are shown.A simplified global approach to creep crack initiation under isothermal conditions, based on reference stress and length concept is developed. A local approach to creep cracking, in which an intergranular physical damage law determined experimentally on notched bars, and stress-strain fields obtained by analytical results is shown to provide crack growth results in good agreement with experiment.     

Fracture mechanics investigation regarding the effects of cracks on the structural integrity of a BWR-core shroud

As a consequence of core shroud intergranular stress corrosion cracking (IGSCC) detected in the course of inservice inspections, a fracture mechanics analysis was carried out to evaluate the effects of postulated cracks on the structural integrity. In this study, critical crack sizes and crack growth were calculated. Due to the comparatively low stress acting on the core shroud during normal operation, the residual stresses in the welds make up the major proportion of the tensile stresses responsible for IGSCC. In order to consider residual stresses of the lower core support ring welds, a finite element analysis was performed at MPA Stuttgart using the FE-code ANSYS. The crack growth computed on the basis of USNRC crack growth rates da/dt demonstrated that crack growth in depth direction increases quickly at first, then retards and finally comes almost to a standstill. The cause of this ‘quasi-standstill’ is the residual stress pattern across the wall, being characterized by tensile stresses in the outer areas of the wall and compressive stresses in the middle of the wall. Crack growth in circumferential direction remains more or less constant after a slow initial phase. As the calculation of stress intensity factors K_I of surface flaws under normal conditions demonstrated, a ‘lower bound’ fracture toughness value is only exceeded in the case of very long and deep surface flaws. It can be inferred from crack growth calculations that under the assumption of intergranular stress corrosion cracking, the occurrence of such deep and at the same time long flaws is unlikely, regardless of the initial crack length. Irrespective of the above, the calculated critical throughwall crack lengths, which were determined using a ‘lower bound’ fracture toughness value, demonstrated that even long throughwall cracks will not affect the component’s integrity under full load. Moreover, it can be concluded from the studies of crack growth that—assuming intergranular stress corrosion cracking—a sufficiently long period will elapse before a crack which has just been initiated reaches a relevant size. Therefore, it can be stated that these cracks will likely be detected during periodic inservice inspections.