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.     

Effect of seawater components on corrosion behavior of SUS316L in nitric acid solution containing metal ions

During the accident that occurred at the Fukushima Daiichi nuclear power plant, a large volume of seawater was introduced as coolant into the storage pools for spent nuclear fuel. If this fuel is reprocessed, some components of seawater will be mixed with the nitric acid solution containing metal ions in the reprocessing process where stainless steels are used as structural material. In this study, we investigated the effect of seawater components in high active liquid waste (HAW) containing nitric acid and metal ions as fission products on the corrosion behavior of SUS316L stainless steel.

Corrosion tests were conducted in surrogate HAW containing artificial seawater (ASW). Intergranular corrosion was observed in the HAW with ASW, where Ru increased the corrosion potential to the transpassive region. An increase in the amount of ASW led to a decrease in the corrosion rate and suppression of intergranular corrosion. Interactions between Ru ions and seawater components, such as chloride ions, were indicated by the results of extended X-ray absorption fine structure spectroscopy and cyclic voltammetry analyses of the solution containing ASW and HAW.     

液态锂铅合金中316L不锈钢的静态腐蚀行为

采用挂片法、失重法和金相分析,开展了结构材料316L不锈钢在液态锂铅(LiPb)合金中静态腐蚀行为的研究.研究结果表明:316L不锈钢中的组分元素,在液态LiPb合金中发生了溶解和质量迁移,这是导致材料腐蚀的主要原因,而温度和合金中的氧含量是影响静态腐蚀行为最重要的参数.     

Corrosion behavior of TiO2-treated type 304 stainless steels in high temperature water containing with hydrogen peroxide

ABSTRACT

Titanium dioxide (TiO₂) treatment coupled with ultraviolet irradiation was selected as a corrosion mitigation technique for Type 304 stainless steel (SS) in high-temperature pure water with the presence of hydrogen peroxide (H₂O₂). Type 304 SS specimens were pre-oxidized in oxygenated pure water at 288 °C and then coated with TiO₂ nanoparticles by hydrothermal deposition. Electrochemical polarization analyses were conducted to investigate the corrosion behavior of both TiO₂-treated and pre-oxidized specimens in 288 °C pure water with 300 ppb H₂O₂. Ultraviolet (UV) irradiation was then imposed upon the TiO₂-treated specimens to examine if there was any distinct photoelectric effect on the corrosion behavior of the treated samples. It was found that the electrochemical corrosion potentials of the TiO₂ treated specimens under UV irradiation were 10–20 mV lower than those without UV. In addition, the corrosion current densities of the treated specimens were also lower in the presence of UV radiation. Without UV radiation, however, no significant differences were observed between the TiO₂ treated and untreated specimens. These results indicate that the TiO₂ treatment in combination with UV radiation would reduce the corrosion rate of Type 304 SS in H₂O₂-rich, high-temperature pure water.     

Effects of gamma-ray irradiation on crevice corrosion repassivation potential of stainless steel in high temperature diluted simulated seawater

The crevice corrosion repassivation potentials (E_R,CREV) of type 304 stainless steel (304 SS) were measured in high temperature (373–553 K), diluted simulated seawater under gamma-ray irradiation, in order to confirm the effects of gamma-ray irradiation on the crevice corrosion behavior of a representative stainless steel in seawater. Overall, for high temperatures, the E_R,CREV values decreased with increasing chloride ion concentration, which was the same as the behavior observed under the non-irradiated condition. The E_R,CREV values measured under gamma-ray irradiation were the same or slightly higher than E_R,CREV values measured under the non-irradiated condition when the [Cl^?] was the same. Consequently, it was confirmed that the threshold potential of crevice corrosion of 304 SS for the gamma-ray irradiation of 1.8 kGy at least did not deteriorate compared with the non-irradiated condition. Under the conditions of this work (seawater composition, [Cl^?] range, dose rate, absorbed dose, flow rate, etc.), the crevice corrosion of 304 SS could be suppressed by maintaining the potential below the threshold potential which was determined approximately as ?0.3 V vs. SHE even for the irradiated condition at temperatures up to 553 K.     

The general corrosion behavior of ENiCrFe-7 weld overlay cladding material in deoxygenated high-temperature and high-pressure water

The general corrosion behavior of Alloy ENiCrFe-7 in deoxygenated high-temperature and high-pressure water was investigated. The results showed that the precipitates of Alloy ENiCrFe-7 included niobium carbide and Al-Ti-O compounds, and the weight gain increased fast firstly before 2250 h, then the weight gain slowed down. There were obvious large particles spread on denser oxide film after 3000 h exposure. Ni was present at a single chemical metallic Ni state, Feⁿ⁺ content of the outer layer was close to 60%, which was much higher than that of the matrix. The oxide film consisted of an inner layer and an outer layer, the inner layer was mainly composed of Cr₂O₃ and the outer layer was mainly composed of Fe₃O₄ and FeCr₂O₄. Finally, it is found that the preferential corrosion location of pitting was niobium carbide precipitates by in same site observation, while Al-Ti-O compounds was not dissolved in deoxygenated high-temperature and high-pressure water for 1500 h exposure. The size and number of the pitting was not significantly changed with increasing exposure time.     

304NG在超临界水中的腐蚀增重随温度的异常关系

研究了奥氏体不锈钢304NG在550、600和650℃超临界水环境下的腐蚀行为。采用扫描显微镜、X射线能谱仪、X射线衍射仪分析了氧化膜的腐蚀形貌、组织结构和成分分布。实验结果表明,试样在3种不同温度下经1000h腐蚀实验后的增重均符合幂函数规律,但650℃时的腐蚀增重与600℃时的相比大幅下降,其主要原因为在较高温时,Cr的扩散速度快,试样表面氧化膜能够维持保护性从而使疖状腐蚀分布数量减少所致。     

Investigation of cobalt buildup behavior and suppression by zinc injection on stainless steel under HWC conditions using simultaneous continuous measurements of corrosion and cobalt buildup

In boiling water reactor (BWR) plants, cobalt-60 (⁶⁰Co) is the main source of radiation exposure, and it builds up on oxide films of structural materials. The ⁶⁰Co buildup is caused by its incorporation into the oxide films. In the BWR plants using hydrogen water chemistry (HWC) to mitigate the oxidative environment, Zn injection has been applied to reduce the ⁶⁰Co incorporation. In this work, we studied the incorporation mechanism of ⁶⁰Co into the oxide films on type 316 stainless steel and the suppression mechanism of ⁶⁰Co incorporation. In order to discriminate between coprecipitation and adsorption of ⁶⁰Co incorporation under HWC conditions, we measured the corrosion amount of the base metal and the ⁶⁰Co buildup amount, using simultaneous continuous measurements for 500 h. The ⁶⁰Co incorporation increased with time both with and without Zn injections. We found that the time dependencies of ⁶⁰Co incorporation with and without Zn have one and two regions, respectively. In the initial stage for both, ⁶⁰Co was incorporated mainly by coprecipitation. After 100 h without Zn, ⁶⁰Co was incorporated by both coprecipitation and adsorption. These results mean that Zn suppressed both coprecipitation and adsorption of ⁶⁰Co.     

核级304L不锈钢钎焊接头组织及耐腐蚀性能研究

核级304L不锈钢与BNi 7钎料真空钎焊接头存在晶间腐蚀行为,但工艺与钎缝耐腐蚀性能的关系尚未得到充分研究。为充分评估压水堆燃料组件结构件中不锈钢真空钎焊接头的晶间腐蚀和应力腐蚀敏感性,降低腐蚀失效风险,采用定量金相方法分析了钎缝中的化合物相含量,采用硫酸 硫酸铁法和双环动电位再活化（DL EPR）法评价了钎缝耐晶间腐蚀性能,并采用高温高压水应力腐蚀裂纹扩展试验评价了钎缝的耐应力腐蚀性能。结果表明,钎缝中化合物相含量越高,耐晶间腐蚀性能越好。且钎缝在高温高压水中存在明显的应力腐蚀开裂行为,但其与钎焊工艺的关系尚需进一步试验研究。     

Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water

The low-frequency corrosion fatigue (CF) crack growth behaviour of different low-alloy reactor pressure vessel steels was characterized under simulated boiling water reactor conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in the temperature range of 240-288 °C with different loading parameters at different electrochemical corrosion potentials (ECPs). Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographical analysis by SEM were used to quantify the cracking response. In this paper the effect of ECP on the CF crack growth behaviour is discussed and compared with the crack growth model of General Electric (GE). The ECP mainly affected the transition from fast (‘high-sulphur’) to slow (‘low-sulphur’) CF crack growth, which appeared as critical frequencies ν_crit = f(ΔK, R, ECP) and ΔK-thresholds ΔK_EAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dt_Air,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔK_EAC-thresholds were shifted to lower values with increasing ECP. The CF crack growth rates of all materials were conservatively covered by the ‘high-sulphur’ CF line of the GE-model for all investigated temperatures and frequencies. Under most system conditions, the model seems to reasonably well predict the experimentally observed parameter trends. Only under highly oxidizing conditions (ECP ? 0 mV_SHE) and slow strain rates/low loading frequencies the GE-model does not conservatively cover the experimentally gathered crack growth rate data. Based on the GE-model and the observed cracking behaviour a simple time-domain superposition-model could be used to develop improved reference CF crack growth curves for codes.     

Effects of seawater components on radiolysis of water at elevated temperature and subsequent integrity of fuel materials

Effects of seawater components on radiolysis of water at elevated temperature have been studied with a radiolysis model and a corrosion test under gamma-ray irradiation conditions to evaluate the subsequent influence on integrity of fuel materials used in an advanced boiling water reactor. In 2011, seawater flowed into the nuclear power plant system of the Hamaoka Nuclear Power Station Reactor No. 5 during the plant shutdown operation. The reactor water temperature was 250 °C and its maximum Cl^? concentration was ca. 450 ppm when seawater was mixed with reactor water. The radiolysis model predicted that the main radiolytic species were hydrogen, oxygen and hydrogen peroxide. Concentrations of radiolytic products originating from Cl^? and other seawater components were found to be rather low. The dominant product among them was ClO₃^? and its concentration was found to be below 0.01 ppm for a 10⁵ s irradiation period. No significant corrosion of zircaloy-2 and 316L stainless steel was found in the corrosion test. These results led to the conclusion that the harmful influence of radiolytic products originating from seawater components on integrity of fuel materials must be smaller than that of Cl^? which is the main ionic species in seawater.     

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。     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.     

Remaining life prediction of the core shroud due to stress corrosion cracking failure in BWRs using numerical simulations

Stress corrosion cracking (SCC) of the welded joints in a reactor core shroud is the primary result of the residual stresses caused by welding, corrosion and neutron irradiation in a boiling water reactor (BWR). Therefore, the evaluation of SCC propagation is important for the safe maintenance of the core shroud. This paper attempts to predict the remaining life of the core shroud due to SCC failures in BWR conditions via SCC propagation time calculations. First, a two-dimensional finite element method model containing H6a girth weld in the core shroud was constructed, and the weld processing was simulated to determine the weld's residual stress distribution. Second, using a basic weld residual stress field, the SCC propagation was simulated using a node release option and the stress redistribution was calculated. Combined with the J-integral method, the stress intensity factors were calculated at depths of 2, 3, 4, 8, 12, 16, 19, 22, 25 and 30 mm in the crack setting inside the core shroud; then, the SCC propagation rates were determined using the relation between the SCC propagation rate and the stress intensity factor. The calculations show that the core shroud could safely remain in service after 9.29 years even when a 1-mm-deep SCC has been detected.     

The Influence of ZrO2 Treatment on the Electrochemical Behavior of Oxygen and Hydrogen on Type 304 Stainless Steels in High Temperature Water

For enhancing the effectiveness of hydrogen water chemistry (HWC) in boiling water reactors (BWRs) in the aspects of lower hydrogen consumption and of a more effective reduction in electrochemical corrosion potential (ECP), the technique of inhibitive protective coating on structural materials was brought into consideration. The application of inhibitive treatment is aimed at deterring the reduction reactions of oxidizing species occurring on metal surfaces and the oxidation reaction of metals. In the current study, electrochemical polarization analyses at 288°C were conducted to characterize the electrochemical properties of ZrO₂ treated and untreated 304 stainless steel specimens in pure water with dissolved oxygen or hydrogen. The polarization results showed that the treated specimens exhibited lower corrosion potentials, corrosion current densities, exchange current densities, and cathodic current densities than the untreated one in high temperature pure water with dissolved oxygen. For the environment with dissolved hydrogen only, reductions in anodic current density and exchange current density were observed, indicating that the ZrO₂ treatment also deterred the oxidation reaction of hydrogen. However, in comparison with the data obtained, the ZrO₂ treatment seemed to be relatively more effective in inhibiting the oxygen reduction reaction than inhibiting the hydrogen oxidation reaction. One additional beneficial outcome was that the anodic current density of the metal was also decreased, leading to a much lower overall corrosion current density of the ZrO₂ treated specimen.     

Measurement and analysis of reflector reactivity worth by replacing stainless steel with zirconium at the fast critical assembly (FCA)

A reflector reactivity worth was measured by replacing stainless steel with zirconium at the FCA. The experimental result of the positive reflector reactivity worth demonstrates the effectiveness of the zirconium reflector compared with the SS reflector in the fast reactor core. This paper also focuses on the validation of standard calculation methods used for fast reactors with JENDL-4.0. As a result, it is confirmed that the standard calculation methods for the reflector reactivity worth show agreement within the experimental error.     

Analyzing the effect of displacement rate on radiation-induced segregation in 304 and 316 stainless steels by examining irradiated EBR-II components and samples irradiated with protons

Recent studies have indicated that, at temperatures relevant to fast reactors and light water reactors, void swelling in austenitic alloys progresses more rapidly when the radiation dose rate is lower. A similar dependency between radiation-induced segregation (RIS) and dose rate is theoretically predicted for pure materials and might also be true in complex engineering alloys. Radiation-induced segregation was measured on 304 and 316 stainless steel, irradiated in the EBR-II reactor at temperatures near 375 °C, to determine if the segregation is a strong function of damage rate. The data taken from samples irradiated in EBR-II is also compared to RIS data generated using proton radiation. Although the operational histories of the reactor irradiated samples are complex, making definitive conclusions difficult, the preponderance of the evidence indicates that radiation-induced segregation in 304 and 316 stainless steels is greater at lower displacement rate.     

Interaction between B4C, 304 stainless steel,and Zircaloy-4 in H2O/H2 atmosphere at 1473 K

The control blade and the fuel rod installed in channel box melt in steam during severe accident of a boiling water reactor. In order to clarify the melting phenomena and relocation of the structural material in the core of the reactor, interaction and melting behavior among B₄C, 304 grade of stainless steel (SST), and Zircaloy-4 in atmosphere containing H₂/H₂O at 1473 K are investigated. The results showed that the reaction at the interface between B₄C and SST under H₂O atmosphere was slow and an oxidation was observed after 3600 s. Under H₂O/H₂ atmosphere, the concentration of B and C in the SST increased and the SST melted. Despite the atmosphere, an oxide layer formed on the surface of Zircaloy-4, and thus the reactions proceeded slowly when the Zircaloy-4 was contacted with B₄C and SST. Under H₂O atmosphere, continuous oxidation happened to SST, and SST was partially melted. Under H₂ atmosphere, the SST was also melted due to the diffusion of B and C from the B₄C. In addition, the oxidation of B₄C affected the oxidation behavior of SST and Zircaloy, and thus the oxidation and the hydroxylation of B₄C in a severe accident was discussed thermodynamically.     

316L不锈钢阻氢或氢同位素渗透搪瓷壁垒层的制备及性能研究

为解决316L不锈钢防氢及其同位素渗透问题,采用两搪两烧方法在不锈钢基体表面制备厚90~110μm的搪瓷涂层。利用X射线衍射仪、光学和扫面电子显微镜表征搪瓷涂层的显微结构、界面形貌,通过EDS线扫描分析界面处主要元素分布,结果显示搪瓷涂层结构致密,与基体形成化学结合,抗落球冲击和热震性能优异。西华特装置气相充氢试验与维氏显微硬度试验结果表明搪瓷涂层是一种有效地阻止氢及其同位素渗透的壁垒层。