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.     

Effects of local radiolysis and geometric parameters on intergranular attack caused by crevice corrosion

Effects of γ-ray irradiation upon crevice corrosion (CC) of type 316L stainless steel (316L SS) as an initiation site of stress corrosion cracking in a boiling water reactor environment have been studied using a material corrosion test loop which could be irradiated with a ⁶⁰Co γ-ray source during testing. The CC tests were conducted using crevice specimens with various sizes of crevice gaps. Many of the examined specimen surfaces exhibited a selective grain boundary dissolution; that is, intergranular attack (IGA) as a result of the CC when the crevice gap size was lower than a certain value. The IGA initiation time was shortened by the γ-ray irradiation. The IGA occurred mostly near the crevice mouth at a distance of less than 2 mm from the mouth edge. When γ-ray exposure had occurred, it was found that the number of IGA sites deeper in the crevice increased compared with the IGA site distribution under the no-irradiation condition. Since the electrochemical corrosion potential inside crevice specimens must be low under the conditions for which IGA could occur, it was assumed that γ-ray irradiation accelerated the corrosion rate of 316L SS by decreasing the Fe²⁺ surface activity inside the crevice or increasing the cathodic current of radiolytic oxidants on the crevice surface. It was concluded that γ-ray irradiation affects the IGA occurrence not only temporally but also spatially.     

Pitting and EPR studies of ion implanted plain and sensitized 304 stainless steel

30 keV N₂⁺, Ti⁺ and N₂⁺ followed by Ti⁺ ions were implanted at room temperature on plain and sensitized 304 stainless steel (304 SS) foils. The ion fluences ranged from 1 × 10¹⁷ ions/cm² to 5 × 10¹⁷ ions/cm². Pitting studies in quiescent non-dearated 3.5 wt.% NaCl electrolyte and Electrolytic Potentiokinetic Reactivation (EPR) studies in a standard electrolyte of KSCN + H₂SO₄ were carried out on plain, sensitized and implanted plain, and sensitized 304 SS foils. In general, the localized corrosion behaviour deteriorated on implantation in plain 304 SS irrespective of the nature of the implant species and their fluences employed in this work. However, the localized corrosion behaviour of implanted sensitized 304 SS showed an improvement. The benefits accrued was dependent on the nature of the implant species and their fluences. An attempt has been made to explain why the implant species interact differently, with the consequent difference in the localized corrosion behaviour, when the stainless steel substrate is in the sensitized state.     

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.     

Evaluation method of corrosion lifetime of conventional stainless steel canister under oceanic air environment

Natural exposure and accelerated corrosion tests of conventional stainless steels for canisters of Types 304, 304L, and 316(LN) for concrete casks were conducted using several test specimens and 1/5 scale canister models. The welding residual stress of a full-scale model canister was also measured and the lifetime of sealability of canisters against corrosion evaluated. The maximum pitting rate and crevice corrosion rate of Type 304 were approximately 20 and 30 μm/year. Many SCC in the 4 Point Bending (4PB) test specimens were found to initiate from the bottom of the corrosion area by pitting or crevice corrosion. The SCC propagation rates in Types 304 and 304L under natural conditions were around 1.2E−12 to 1.8E−11 m/s in the K (Stress Intensity Factor) range of 0.6–9.0 MPa m^1/2, and that of the accelerated test (60 °C, 95% RHS, filled with NaCl mist) around 1.0E−10 to 3.5E−9 m/s in the K range of 0.5–30 MPa m^1/2. The SCC propagation rates under both natural and accelerated conditions were independent of K. The lifetime of sealability estimated from 1/5 scale models was longer than that from the small bending test specimens and has a safety margin as a structure.     

Cathodic polarization curves of the oxygen reduction reaction on various structural materials of boiling water reactors in high temperature–high purity water

Cathodic polarization curves of the O₂ reduction reaction were measured by using electrodes made from typical structural materials of boiling water reactors (BWRs) to evaluate the effects of kind of material on the electrochemical corrosion potential (ECP) calculation. To estimate ECPs at any region in the BWRs on the basis of the BWR environmental conditions, anodic and cathodic polarization curves should be obtained in advance under relevant conditions. The concentration of oxidants such as O₂ and H₂O₂ in coolant changes depending on the region in which they exist. As well, reduction reaction rates might differ depending on the kind of materials. In this work, the cathodic polarization curves of type 316L stainless steel (316L SS) and Alloy 182 were measured in high purity water at 553 K with different O₂ concentrations and compared with those of type 304 SS (304 SS). The results showed that the cathodic polarization curves differed depending on the kind of materials at the activation-controlled region. But, the difference in the ECP vs. O₂ concentration relationship was small when the ECPs were calculated by using both anodic and cathodic polarization curves measured on the objective material.     

Water Chemistry in a Crack Tip under Irradiation, (II)

Under neutron and gamma-ray irradiations, radiolytic species are generated directly in the crack tip, which causes higher oxidant concentrations and subsequently influences crack propagation rate.

A crevice radiolysis model was proposed to estimate the oxidant concentrations in the crack tip water under gamma-ray irradiation. Direct generation of radiolytic species in the crevice water, and their secondary generation and disappearance caused by their interaction with the crevice surface as well as species in the crevice water were included in the model. The diffusion of the radiolytic species through the narrow gap from the bulk water to the crack tip and vice versa were also considered.

Calculation results confirmed that the concentrations of H₂O₂, one of the most important oxidants in BWR environments, in both bulk water and crack tip water under irradiation (energy deposition rate: 0.1 W/cm) were high enough to show high local ECP in both regions under NWC, but were high in the bulk water and low in the crack tip water under HWC. A high H₂ diffusion rate from the bulk to the crack tip enhanced the recombination reaction of H₂O₂ and H₂.     

Gamma-ray Irradiation Effects on Corrosion Rates of Stainless Steel in Boiling Nitric Acid Containing Ionic Additives

Irradiation effects of γ-rays on corrosion rates of type 304ULC stainless steel in 9×10^?3 mol/m³ boiling nitric acid containing an ionic additive of multivalence elements of Ce(IV), Cr(VI) or Ru(III) were studied by measuring weight losses of specimens immersed under the ⁶⁰Co γ-ray irradiation of 1 kC/(kg·h) (4MR/h). Tests without irradiation were carried out as well to obtain reference data. All the coexisting ionic species enhanced the corrosion in comparison with those in pure nitric acid, and the γ-ray irradiation moderated the enhancement. The valence analyses of these additives and redox potential measurements before and after the immersion batch under the irradiation suggested that the reduction of Ce(IV) and Cr(VI) induced by the irradiation have moderated the corrosion environment.     

Intergranular stress corrosion cracking susceptibility of sensitized type 304 steel in 561 K pure water under γ-ray irradiation

Intergranular stress corrosion cracking (IGSCC) of sensitized type 304 stainless steel has been investigated in 561 K water under γ-ray irradiation at a flux of 2.6 × 10³ C kg⁻¹ h⁻¹ by slow-strain-rate tensile tests. The IGSCC susceptibility was enhanced by γ-ray irradiation in water containing 8 ppm dissolved oxygen (DO). The DO dependence of the IGSCC susceptibility was observed in the water under γ-irradiation. Although slight IGSCC susceptibility was observed even in deaerated water (less than 1 ppb DO) under γ-ray irradiation, the susceptibility was completely suppressed by injection of hydrogen into the water. The enhancement of IGSCC susceptibility seems to be related to the formation of H₂O₂ in high temperature water by radiolysis under γ-ray irradiation and the H₂O₂ formation rate is markedly decreased by hydrogen injection.     

Determining factors for anodic polarization curves of typical structural materials of boiling water reactors in high temperature – high purity water

In order to examine the anodic polarization characteristics of typical structural materials of boiling water reactors (BWRs), the anodic polarization curves of type 316L stainless steel (316L SS) and Alloy 182 were measured in deaerated high purity water at 553 K using the previously reported measurement method which was confirmed suitable for high temperature – high purity water. In order to specify which constituent element determines the dissolution characteristics of these materials, the anodic polarization curves of pure iron, pure nickel, and pure chromium were also surveyed. The anodic polarization curve of 316L SS was determined to have active, passive, and transpassive states which were the same as type 304 SS (304 SS) showed. But, Alloy 182 had different polarization characteristics especially near the corrosion potential as it had no active state. From comparison results of the polarization characteristics of these materials and their constituent elements, the corrosion characteristics of these materials were concluded to be mainly determined by the corrosion characteristics of chromium.     

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

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

Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 10²⁵ n/m² (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σ_x residual stress (transverse to the weld bead) and σ_y residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10⁻⁶/MPa/dpa). This value was smaller than that of nickel alloy.     

核级不锈钢焊接接头钝化膜耐蚀性能和半导体特性研究

采用电化学阻抗谱(EIS)、场发射扫描电子显微镜(SEM)、Auger扫描能谱仪(AES)以及容抗测试技术(M-S曲线),研究了316LN/316L不锈钢焊接接头在模拟压水堆一回路高温高压水中形成的钝化膜的耐蚀性能和半导体特性。结果表明,焊缝区、热影响区和母材区形成的钝化膜的耐蚀性能不同,热影响区钝化膜开路电位及电化学阻抗等均低于其他区域,说明热影响区钝化膜的耐蚀性能最差,这主要与钝化膜的致密程度、厚度及Cr氧化物的含量有关。M-S曲线表明,母材区钝化膜平带电位为-0.7V,较其他区域(-0.4V)负移,表明有BO-3等阴离子在钝化膜表面吸附,加之具有较低的施主和受主浓度,可排斥侵蚀离子的腐蚀,使之较其他区域有更强的耐蚀性能。     

Effect of prior thermal treatment on the microchemistry and crack propagation of proton-irradiated AISI 304 stainless steels

The effect of prior thermal treatment on crack growth was investigated on proton-irradiated Type 304 stainless steel (SS) of initially solution annealed (SA) and thermally sensitized (SEN) conditions. The Cr depletion profiles were measured by field emission gun transmission electron microscopy/energy dispersive spectroscopy (FEGTEM/EDS) in an attempt to correlate grain boundary chromium composition with the measured crack growth rate. The results showed that the crack growth of the 1-dpa-irradiated SEN 304SS is substantially higher than that of SA 304SS with the same irradiation dose. The unirradiated SEN material initially started with a shallow Cr depletion profile near grain boundary. After 1 dpa irradiation with proton, the Cr depletion profile becomes narrower and deeper. In contrast, the grain boundary Cr concentration in the SA specimen at the same irradiation dose was higher than that of the SEN specimen, mainly due to an initial Cr enriched condition. Consequently, the irradiated SEN specimen exhibited a higher degree of sensitization in electrochemical potentiokinetic reactivation test and faster crack growth rate in the stress corrosion crack test. The absence of irradiation enhanced crack growth in heavily thermal-sensitized 304SS is probably attributed to slower radiation-induced Cr depletion as a result of pre-existing thermally induced grain boundary Cr depletion. It is a clear indication that the inverse Kirkendall effect was hampered by the back diffusion of Cr due to initially depleted Cr concentration gradient near grain boundary.     

Residual stress measurements in irradiated solution-annealed type 304 stainless steel tubing

Immersion density and residual stress measurements were made on solution-annealed type 304 stainless steel capsule tubing irradiated up to fluence levels of 9 × 10²² n/cm² (E > 0.1 MeV). The measured residual stress is dependent on the competition between differential swelling which builds up differential stresses, and irradiation creep which relaxes these stresses. The measurements were analyzed using a bilinear swelling equation formulated with swelling data from the same heat of material. The temperatures and fluence levels of the swelling and slit tube data were each calculated with the same computer code. At high fluence, when swelling was in the steady-state region, the effective irradiation creep rate increased by a factor of about three. Further analysis was made assuming that stress-enhanced swelling and swelling-enhanced irradiation creep were the enhanced relaxation mechanisms.     

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.     

The effect of ionizing radiation on the corrosion behavior of metals in carbon tetrachloride

It has been found that ionizing radiation considerably intensifies the corrosion of a number of metals in moist CCl⁴. The intensification of the corrosion process is due to the presence of hydrochloric acid and oxidizing agents — energetic cathodic depolarizers, forming during the radiolysis of moist CCl₄.Increasing the amount of moisture above its solubility limits in CCl₄ leads to a weakening of the corrosion process because of the diffusion of Cl^– into the aqueous phase.A detailed study has been made of the mechanism of formation of HCl in the studied system under irradiation and the effect of the composition of the atmosphere on the corrosion process.     

Irradiation induced stress-relaxation of previously irradiated 304 stainless steel in a fast flux environment

Results of a recent fast flux neutron irradiation experiment in EBR-II designed to determine the effects of high levels of prior irradiation (to 10²³ n/cm², $\text{E}$ > 0.1 MeV) on the irradiation creep of type 304 stainless steel at 800° F are reported. The primary conclusion drawn from the data is that the steady state creep coefficient increases by a factor of 8 as the specimen fluence increases from 0 to 10.0 × 10²² n/cm² ($\text{E}$ > 0.1 MeV). The irradiation creep coefficients are consistent with a linear variation in creep rate with swelling rates over the entire data range. The restrictions that the experimental results place on the choice of a theoretical model for irradiation creep are cited.     

Corrosion of ZrO2 treated type 304 stainless steels in high temperature pure water with various amounts of hydrogen peroxide

As boiling water reactors (BWRs) age, intergranular stress corrosion cracking (IGSCC) of the structural materials in the reactor piping systems and vessel internals has become a major degradation problem. Several approaches to mitigating IGSCC in the structural components have been developed and investigated. Among them, the technique of inhibitive protective coatings is deemed the most promising one since it is expected to work even in the absence of the well-known hydrogen water chemistry technology.Following our earlier work on exploring the electrochemical characteristics of important oxidizing species on zirconium oxide (ZrO₂) treated Type 304 stainless steels (SSs), we targeted on the characteristics of hydrogen peroxide, which is another strongly oxidizing species in the reactor coolant other than oxygen, in this study. Tests were conducted to determine electrochemical parameters such as electrochemical corrosion potential (ECP), corrosion current density, exchange current density and Tafel constant of the reduction reaction of hydrogen peroxide on 304 SS specimens before and after the ZrO₂ treatment. The surface morphologies of the treated and untreated specimens were examined by scanning electron microscopy, energy dispersive X-ray spectroscopy, and laser Raman spectra. Furthermore, the corrosion mitigation efficiency of ZrO₂ treatment was evaluated by electrochemical polarization tests in simulated BWR environments. Test results showed that there were no significant differences in ECP between the untreated and ZrO₂ treated specimens in the test environments of various hydrogen peroxide concentrations. However, it was found via polarization analysis that the exchange current density of the reduction reaction on and the corrosion current density of the treated specimens were markedly lower than those on and of the untreated ones in the same environments. The ZrO₂ treatment was able to deter the reduction rate of hydrogen peroxide on the Type 304 SS surface.     

Irradiation creep of previously irradiated 304 stainless steel in a fast flux environment

Results of a fast flux neutron irradiation experiment designed to investigate the effects of high levels of prior irradiation (to 10²³ n/cm², E > 0.1 MeV) on the irradiation creep of type 304 stainless steel at 700 K (800°F) are reported. The steady state creep coefficient is found to increase by a factor of 7 as the specimen fluence increases from 3 to 10 × 10²²n/cm², (E > 0.1 MeV). A non-linear dependence of strain on stress is exhibited. The results of this study are compared with previously reported stress relaxation data and with predictions of a swelling enhanced irradiation creep model.