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1.
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 KI60 MPa m1/2). 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 KIJ or with a high degree of plasticity in the remaining ligament. 相似文献
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Hiromu Ambai Yusuke Nishizuka Yuichi Sano Naoki Uchida Shizuka Iijima Hideaki Shiwaku 《Journal of Nuclear Science and Technology》2019,56(2):193-200
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. 相似文献
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ABSTRACTTitanium 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. 相似文献
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Masahiko Tachibana Kazushige Ishida Yoichi Wada Nobuyuki Ota Naoto Shigenaka 《Journal of Nuclear Science and Technology》2016,53(7):981-991
The crevice corrosion repassivation potentials (ER,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 ER,CREV values decreased with increasing chloride ion concentration, which was the same as the behavior observed under the non-irradiated condition. The ER,CREV values measured under gamma-ray irradiation were the same or slightly higher than ER,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. 相似文献
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Peiliang Guo Ping Zhu Xinyuan Cao Wei Wang Ligong Ling 《Journal of Nuclear Science and Technology》2019,56(4):355-363
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. 相似文献
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Toshimasa Ohashi Tsuyoshi Ito Hideyuki Hosokawa Makoto Nagase Motohiro Aizawa 《Journal of Nuclear Science and Technology》2013,50(4):588-595
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. 相似文献
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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。 相似文献
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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 νcrit = f(ΔK, R, ECP) and ΔK-thresholds ΔKEAC = f(ν, R, ECP) in the cycle-based form and as a critical air fatigue crack growth rate da/dtAir,crit in the time-domain form. The critical crack growth rates, frequencies, and ΔKEAC-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 mVSHE) 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. 相似文献
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Yoichi Wada Masahiko Tachibana Atsushi Watanabe Kazushige Ishida Nobuyuki Ota Naoto Shigenaka 《Journal of Nuclear Science and Technology》2016,53(6):809-820
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. 相似文献
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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. 相似文献
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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. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(9):809-815
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. 相似文献
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Masahiro Fukushima Yasunori Kitamura Masaki Andoh Teruhiko Kugo 《Journal of Nuclear Science and Technology》2013,50(10):961-965
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. 相似文献
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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. 相似文献
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Xu Gao Ryosuke Sasaki Lichun Zheng Sun-joong Kim Shin-ya Kitamura 《Journal of Nuclear Science and Technology》2018,55(4):400-409
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. 相似文献