Stress corrosion cracking on irradiated 316 stainless steel

Tests on irradiation-assisted stress corrosion cracking (IASCC) were carried out by using cold-worked (CW) 316 stainless steel (SS) in-core flux thimble tubes which were irradiated up to 5×10²⁶ n/m² (E>0.1 MeV) at 310°C in a Japanese PWR. Unirradiated thimble tube was also tested for comparison with irradiated tubes. Mechanical tests such as the tensile, hardness tests and metallographic observations were performed. The susceptibility to SCC was examined by the slow strain rate test (SSRT) under PWR primary water chemistry condition and compositional analysis on the grain boundary segregation was made. Significant changes in the mechanical properties due to irradiation such as a remarkable increase of strength and hardness, and a considerable reduction of elongation were seen. SSRT results revealed that the intergranular fracture ratio (%IGSCC) increased as dissolved hydrogen (DH) increased. In addition, SSRT results in argon gas atmosphere showed a small amount of intergranular cracking. The depletion of Fe, Cr, Mo and the enrichment of Ni and Si were observed in microchemical analyses on the grain boundary.     

Emulation of neutron irradiation effects with protons: validation of principle

This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 °C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 °C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 °C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a ‘W' shaped profile at 1.0 dpa and then into a ‘V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary denuded zones were only observed in neutron-irradiated samples. No cavities were observed for either irradiating particle. For both irradiating particles, hardening increased with dose for both heats, showing a more rapid increase and approach to saturation for heat B. In normal oxygenated water chemistry (NWC) at 288 °C, stress corrosion cracking in the 304 alloy was first observed at about 1.0 dpa and increased with dose. The 316 alloy was remarkably resistant to IASCC for both particle types. In hydrogen treated, de-oxygenated water (HWC), proton-irradiated samples of the 304 alloy exhibited IG cracking at 1.0 dpa compared to about 3.0 dpa for neutron-irradiated samples, although differences in specimen geometry, test condition and test duration can account for this difference. Cracking in heat P in HWC occurred at about 5.0 dpa for both irradiating particles. Thus, in all aspects of radiation effects, including grain boundary microchemistry, dislocation loop microstructure, radiation hardening and SCC behavior, proton-irradiation results were in good agreement with neutron-irradiation results, providing validation of the premise that the totality of neutron-irradiation effects can be emulated by proton irradiation of appropriate energy.     

FeCrAl铁素体不锈钢在压水堆水化学偏离工况下的应力腐蚀行为

采用慢应变速率拉伸（SSRT）试验方法,研究了FeCrAl铁素体不锈钢在偏离压水堆正常工况下的应力腐蚀行为。结果表明,在含有微量Cu²⁺与Cl^-的高温水介质中,应变速率为2×10^-7s^-1时,FeCrAl铁素体不锈钢将发生显著的应力腐蚀开裂;其在含Cu²⁺和Cl^-的高温水介质中的应力腐蚀开裂主要由点蚀导致。     

Role of Radiation-Induced Grain Boundary Segregation in Irradiation Assisted Stress Corrosion Cracking

Isolation of microstructural and microchemical effects on irradiation assisted stress corrosion cracking (IASCC) was attempted by means of low-dose high-temperature neutron irradiation in a material test reactor to get better understanding on IASCC. Microstructure, grain boundary segregation, hardness and SCC susceptibility were examined on stainless steels irradiated to 0.8 dpa at around 673 K. The irradiation caused well-developed grain boundary segregation without notable hardening or microstructural changes. No IASCC was found in 593 K hydrogenated water whereas the steels were highly susceptible to IASCC in 561 K oxygenated water. The results suggest that grain boundary segregation, probably Cr depletion, is sufficient to cause IASCC in oxygenated water and that other radiation-induced changes such as microstructure and hardening are required for IASCC in hydrogenated water.     

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.     

A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels

Austenitic stainless steels (SSs) are used extensively as structural alloys in the internal components of light water reactor (LWR) pressure vessels because of their relatively high strength, ductility, and fracture toughness. However, exposure to neutron irradiation for extended periods changes the microstructure (radiation hardening) and microchemistry (radiation-induced segregation) of these steels, and degrades their fracture properties. Irradiation-assisted stress corrosion cracking (IASCC) is another degradation process that affects LWR internal components exposed to neutron radiation. The existing data on irradiated austenitic SSs were reviewed to evaluate the effects of key parameters such as material composition, irradiation dose, and water chemistry on IASCC susceptibility and crack growth rates of these materials in LWR environments. The significance of microstructural and microchemistry changes in the material on IASCC susceptibility is also discussed. The results are used to determine (a) the threshold fluence for IASCC and (b) the disposition curves for cyclic and IASCC growth rates for irradiated SSs in LWR environments.     

Effects of Dissolved Hydrogen and Strain Rate on IASCC Behavior in Highly Irradiated Stainless Steels

IASCC behavior in cold-worked SUS316 stainless steels irradiated to 35 dpa was examined using slow strain rate tensile testing at a strain rate of 6.7 × 10^?8/s in 320°C simulated PWR primary water while varying the dissolved hydrogen (DH) concentration from 0 to 2.8 ppm. The results were compared with those previously obtained at a higher strain rate using specimens of different sizes and with those of the previous interrupted experiment. The initiation and propagation of IASCC enhanced with increasing DH concentration and lower strain rate. The IASCC initiation stress decreased to almost half of the yield strength at high DH. Accompanying slow tensile tests in an argon gas environment showed that a lower strain rate did not change in the initiation stress that exceeded the yield strength, but enhanced the propagation of intergranular cracking.     

高温水中氯离子对316L不锈钢应力腐蚀裂纹扩展速率的影响

本文采用直流电压降(DCPD)方法,使用恒K(K=27.5 MPa·m1/2)加载方式,在核电厂高温高压水环境中研究了氯离子对316L不锈钢的应力腐蚀裂纹扩展速率的影响。实验结果表明:在高温除氧水中,氯离子会加快316L不锈钢的应力腐蚀裂纹扩展速率,且当水中存在溶解氧时,氯离子对应力腐蚀裂纹扩展速率的影响更明显。     

Materials aspects of BWR plant life extension

Considerable experience with plant equipment performance in nuclear power stations has indicated that the principal factors limiting the life of BWRs and PWRs are materials related. Specifically, for LWRs it is known that these materials issues generally include parameters related to stress corrosion cracking, corrosion fatigue, wear and radiation embrittlement. Not only do these parameters affect and limit the actual useful design life of plant components but also affect the plant's operating availability. In all these cases, the elimination or control of one or more of these critical parameters should improve the plants availability and significantly extend the useful service life.In the present paper, research performed to address the intergranular stress corrosion cracking (IGSCC) area is described. Specific emphasis is placed on Type 304 stainless steel which has suffered IGSCC in piping in the heat-affected-zone (HAZs) adjacent to the welds in the BWR primary system. Research has developed and qualified a number of techniques which address the three necessary conditions for IGSCC in BWRs: (1) sensitized microstructure, i.e., chromium depletion at the grain boundaries during welding; (2) over yield tensile stress; and (3) oxygenated (200 ppb) high temperature (288Another potential life-limiting IGSCC phenomenon for certain components, irradiation assisted stress corrosion cracking (IASCC) of stainless steel exposed to a high neutron flux, is also discussed. Unlike the IGSCC, IASCC results in intergranular cracking of annealed material at low stress. Fortunately, preliminary research has indicated that some of the techniques utilized for IGSCC control in piping as well as new controlled impurity level stainless steel alloys may reduce the future potential IASCC concern to an insignificant level.     

Sustained load crack growth in A533B-1 steel under neutron irradiation in a water environment

A study of the combined effects of radiation, water and temperature on sustained load crack growth behavior of reactor pressure vessel steel A533B-1 is reported. To complete this study wedge opening loading (WOL) T-type fracture toughness specimens were prepared from a sample of A533B-1 steel which had a copper content of 0.13%. The crack length change was measured after 939 hr of irradiation in a water environment. An electrical potential method was successfully used to measure the crack length of rusted radioactive specimens. Sustained load crack growth occurred at initial stress intensity factor K_Ii as low as . The value of stress corrosion cracking threshold factor K_Iscc after neutron irradiation in a water environment appears to be in the range of . The results of neutron irradiation in a water environment are to apparently increase the susceptibility of A533B-1 steel to stress corrosion cracking and hydrogen embrittlement.     

Analysis of water radiolysis in relation to stress corrosion cracking of stainless steel at high temperatures - Effect of water radiolysis on limiting current densities of anodic and cathodic reactions under irradiation

Electrochemical corrosion potential (ECP) is an important measure for environmental factor in relation to stress corrosion cracking (SCC) of metal materials. In the case of SCC for in-core materials in nuclear reactors, radiolysis of coolant water decisively controls ECP of metal materials under irradiation. In the previous models for ECP evaluation of stainless steel, radiolysis of reactor water in bulk was considered to calculate the bulk concentrations of the radiolysis products. In this work, the radiolysis not only in bulk but also in the diffusion layer at the interface between stainless steel and bulk water was taken into account in the evaluation of ECP. The calculation results shows that the radiolysis in the diffusion layer give significant effects on the limiting current densities of the redox reactions of the radiolysis products, H₂O₂ and H₂, depending on dose rate, flow rate and water chemistry, and leads to the significant increase in the ECP values in some cases, especially in hydrogen water chemistry conditions.     

Diagnostics of core barrel vibrations by in-core and ex-core neutron noise

Diagnostics of core-barrel vibrations has traditionally been made by use of ex-vessel neutron detector signals. We suggest that in addition to the ex-core noise, also the in-core noise, induced by core barrel vibrations, be also used. This would enhance the possibilities of diagnostics where the number of the ex-core detectors is not sufficient or their positions are disadvantageous for effective diagnostics, especially for shell-mode vibrations.

To this order, the theory of in-core noise induced by a fluctuating core boundary has been elaborated and applied to the diagnostics of beam and shell mode vibrations. The formulas were tested on some measurements taken in the Ringhals PWRs. The results confirm the validity of the model itself, and the possibilities for enhanced diagnostics were demonstrated. A more effective use of these novel possibilities requires more in-core detectors and/or better detector positioning.     

焊接残余应力对316LN不锈钢应力腐蚀裂纹扩展速率的影响

核电站不锈钢管道焊接过程中引入的残余应力对焊接接头的应力腐蚀开裂性能有较大影响。本文针对一AP1000主管道316LN不锈钢焊接模拟件进行残余应力分析和应力腐蚀裂纹扩展速率测量,得到了焊后原始状态和去应力热处理状态的焊接热影响区材料在高温高压水中的应力腐蚀裂纹扩展速率。实验结果表明,焊接残余应力明显提高了热影响区的应力腐蚀裂纹扩展速率,且在含氢的压水堆一回路正常水化学下焊接残余应力的影响更加显著。     

316LN在模拟压水堆一回路异常水化学条件下的应力腐蚀敏感性研究

采用慢应变速率拉伸试验（SSRT）法研究了氯离子(Cl^-)和溶解氧对316LN在模拟压水堆（PWR）一回路异常工况下的应力腐蚀开裂（SCC）的影响。结果表明：在饱和氧条件下,随着Cl^-浓度的增大,试样的延伸率降低,SCC敏感指数增大,当Cl^-浓度大于1 mg/L后开始萌生SCC裂纹;溶液除氧后,即使Cl^-浓度增大到10 mg/L也不会引发SCC裂纹。316LN的SCC裂纹萌生于表面的缺陷或贫Cr区,呈穿晶型向基体内扩展。     

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

The effect of neutron irradiation on the iodine stress corrosion cracking (SCC) of Zircaloy-2 tubing of 8×8 type design was studied by means of ring tension test, using specimens unirradiated and irradiated to 3.2×l0¹⁹ and 3.0×10²⁰ n/cm² (E>lMeV). The SCC threshold stresses were determined from constant load tests and the SCC initiation stresses were defined from constant cross-head rate tests. Both stresses increased with increasing neutron fluence, reaching a maximum at a neutron fluence between 10¹⁹ and 10²⁰ n/cm² and then decreased. The tendency is qualitatively in good agreement with the hypothetical conclusion derived by Lunde & Videm, for SCC failure stresses from internal gas pressurization tests on various Zircaloy cladding tubes irradiated at different reactor conditions. The cause of the increase in the SCC susceptibility at neutron fluences above 10²⁰ n/cm² would be ascribed to radiation anneal hardening phenomenon and resultant inhomogeneous incipient deformation characterized by dislocation channels.     

Recovery of Heat-Generating Nuclides in Eluted Solutions

To study about microstructure and chemical composition of oxide films formed on surface of stainless steel is most important for understanding of stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC). In this work, a new sample preparation method for microstructure observation of oxide films was developed. To prevent to break oxide films during fabrication, surface of specimens were protected with plating. Focused ion beam (FIB) processing was conducted to prepare thin foil samples of cross section of oxide films. After sample preparation, microstructure of cross section of oxide films was observed by transmission electron microscope (FE-TEM), and microscopic chemical composition was analyzed by energy dispersed X-ray spectrometer (EDS). From the results, effects of silicon (Si) doping for oxide film formation in two oxidation conditions are discussed.     

Trace elements in reactor steels: implications for decommissioning

Trace elements in stainless steel have been systematically examined for the production of long-lived radioisotopes through neutron activation in reactors. Niobium-94 has been identified as the most important impurity. It is a long-lived ( yr) gamma ray emitter (0.7 and 0.87 MeV), which is produced by the neutron capture reaction ⁹³Nb(n, γ)⁹⁴Nb. Through X-ray fluorescence Nb concentrations of 160 ± 20 ppm have been found in type 304 stainless steel which agrees with earlier published values. At this concentration, the gamma radiation dose rate inside the pressure vessel of a reactor would remain close to 1 rem/hr for thousands of years after the ⁶⁰Co activity has decayed. This could be important for the delayed dismantling option considered for reactors.Nitrogen as an impurity in stainless steel has been shown to result in the buildup of 100 and 1000 Ci of carbon-14 over the lifetime of a BWR and a PWR, respectively. Although ¹⁴C is only a β⁻-emitter, its long half-life (5730 yr) and the crucial role of carbon in the biosphere may be important in deciding on the ultimate disposal method of the radioactive reactor components.     

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.     

Identification of reactor internals'' vibration modes of a Korean standard PWR using structural modeling and neutron noise analysis

The vibration characteristics of a Korean standard PWR reactor internals have been estimated through a three-dimensional finite element analyses and verified by using the mode separated power spectral density functions obtained from the ex-core neutron noise signals. Also the natural vibration modes of the fuel assembly have been identified measuring both the ex-core and the in-core neutron noise signals which are close to each other. As a result, the fundamental bending mode frequency of the reactor internal structure is found to be around 8 Hz and the fundamental shell mode frequency 14.5 Hz, respectively. It is also shown that the fundamental bending mode frequency of the fuel assembly is 2.3 Hz and the 2^nd bending mode frequency 5.8 Hz, respectively. These results can be used for the supplements of the Korean standard PWR's CVAP (Comprehensive Vibration Assessment Program) data.     

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.