Takashi Tsukada

Effects of Hydrogen Peroxide on Corrosion of Stainless Steel, (V) Characterization of Oxide Film with Multilateral Surface Analyses

In order to understand corrosion behavior of stainless steel in BWR reactor water conditions, characteristics of oxide films on stainless steel specimens exposed to H2O2 and O2 in high temperature water were determined by multilateral surface analyses, i.e., SEM (scanning electron microscope), LRS (laser Raman spectroscope), SIMS (secondary ion mass spectroscope) and STEM-EDX (scanning transmission electron microscope). The following points were experimentally confirmed. 1. Oxide layers were divided into inner and outer layers: Outer layers of the specimen exposed to 100ppb H2O2 consisted of larger corundum type hematite (α-Fe2O3) particles, while inner layers consisted of very fine Ni rich magnetite (Fe3O4). Outer layers of the specimen exposed to 200 ppb O2 consisted of larger magnetite mixture particles, while inner layers consisted of fine Cr rich magnetite. 2. Outer oxide layers consisted of oxide particles. The oxide particles depositing on the specimens exposed to 100ppb H2O2 were divided into two groups, i.e., a larger particle group and a smaller particle group. For other specimens, the diameter distribution of depositing particles was a single peak. Particle density and size were changed by oxidant concentration. The average diameter of the particles (that of the smaller group only for the specimen expose to 100 ppb H2O2) decreased with [O2] and [H2O2]. 3. Total oxide film thickness decreased with [H2O2] and increased with [O2]. 4. A larger dissolution rate at higher [H2O2] resulted in a thinner oxide film with smaller particles and larger hematite particles.

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.

Effect of gamma radiolysis on pit initiation of zircaloy-2 in water containing sea salt

In spent fuel pools at the Fukushima Daiichi Nuclear Power Station (1F), seawater was injected for cooling purposes after the tsunami disaster in March 2011. It is well known that the chloride in the seawater has the potential to cause localized corrosion (e.g., pitting corrosion) in metals. In this study, we evaluated the pitting potentials of zircaloy-2, the material used in the fuel cladding tubes in 1F, as a function of chloride concentration. To accomplish this, we used artificial seawater under gamma-ray irradiation and investigated the effect of radiolysis on pit initiation of zircaloy-2 in water containing sea salt. Changes in the composition of water containing sea salt were analyzed as well, both before and after gamma-ray irradiation. The characteristics of the resultant oxide films formed on zircaloy-2 were evaluated by X-ray photoelectron spectroscopy and electrochemical impedance spectroscopy. The experimental results showed that the pitting potential under irradiation was slightly higher than that under conditions in which no radiation was present, and that the pitting potential decreased with increasing chloride concentration in the presence as well as the absence of radiation. Solution analyses for water containing sea salt showed that hydrogen peroxide was generated by irradiation. The oxide film was composed of zirconium oxide and was made thicker during the irradiation. The higher pitting potential could thus be explained by the capacity of hydrogen peroxide to oxidize the surface and enhance oxide film formation. Under gamma-ray irradiation, the zircaloy-2 surface with an oxide film formed by radiolysis products was found to be resistant to pitting in the presence of chloride.

Determination of Electrochemical Corrosion Potential Along the JMTR In-Pile Loop - I: Evaluation of ECP of Stainless Steel in High-Temperature Water as a Function of Oxidant Concentrations and Exposure Time

In-pile loop experiments are one of the key technologies that can provide an understanding of corrosion behaviors of structural materials in nuclear power plants (NPPs). The experiments should be supported not only by reliable measurement tools to confirm corrosive conditions under neutron and gamma-ray irradiations but also by theoretical models for extrapolating the measured data to predict corrosion behaviors in NPPs. The relationships among electrochemical corrosion potential (ECP), metal surface conditions, exposure time, and other environmental conditions have been determined from in situ measurements of corrosion behaviors of stainless steel specimens exposed to H2O2 and O2 in high-temperature water. Based on the relationships, a model to evaluate the ECP of stainless steel was developed by coupling an electrochemical model and a double-oxide layer model. Major conclusions obtained from the evaluation model are as follows: (a) The difference in ECP behaviors of the specimens exposed to H2O2 and O2 were mainly from the thickness and developing rate of the inner oxide layers. (b) Calculated ECP behaviors, e.g., the different responses to H2O2 and O2 and hysteresis and memory effects, agreed with the measured ones. (c) Neutron exposure might decrease ECP due to radiation-induced diffusion in the oxide layer. The ECP evaluation model will be applied to evaluation of corrosive conditions in the JMTR in-pile loop.

Hydrogen Peroxide Production by Gamma Radiolysis of Sodium Chloride Solutions Containing a Small Amount of Bromide Ion

Abstract Gamma radiolysis experiments on solutions of a mixture of sodium chloride (NaCl) and sodium bromide (NaBr) were conducted to confirm the validity of radiolysis calculations for simulated seawater solutions and to determine the importance of bromide anion (Br−) in the production of hydrogen peroxide (H2O2) via water radiolysis. The H2O2 concentration in each solution was measured after irradiation and compared with that obtained from radiolysis calculations. It was found that the calculated and experimental results were in good agreement. The concentration of H2O2 in a 0.6 M NaCl solution increased approximately three times on the addition of 1 mM NaBr. The result showed that Br− plays an important role in the production of H2O2 by water radiolysis, presumably through the reactions of Br− with hydroxyl radical (•OH). For 1 mM NaCl solutions, there is a minimum production rate of H2O2 at pH 8, which increases when the pH changes to either lower or higher values. It was considered that the hydrated electron also plays an important role in H2O2 production under these acidic and alkaline conditions.

Effects of hydrazine addition and N2 atmosphere on the corrosion of reactor vessel steels in diluted seawater under gamma-rays irradiation

Seawater was injected into the reactor cores in the Fukushima Daiichi Nuclear Power Station. Corrosion of primary containment vessel (PCV) steel and reactor pressure vessel (RPV) steel is considered to progress until the molten fuel debris is removed. To evaluate durability of the PCV and RPV steels, corrosion tests were conducted in diluted seawater at 50 °C under gamma-rays irradiation of dose rates of 4.4 and 0.2 kGy/h. To evaluate the effect of hydrazine (N2H4) as an oxygen scavenger under gamma-rays irradiation, 10 and 100 mg/L N2H4 were added to the diluted seawater. Without addition of N2H4, weight loss in the PCV and RPV steels irradiated with the 0.2 kGy/h dose rate was comparable with those without irradiation and weight loss in the vessel steels irradiated with the 4.4 kGy/h dose rate was higher than those without irradiation. Under irradiation, weight loss in the PCV and RPV steels in diluted seawater containing N2H4 was comparable with that in diluted seawater without N2H4. When gas phase in the flask was replaced with N2, weight loss in the PCV and RPV steels, and O2 and H2O2 concentrations in the diluted seawater decreased.

Interrelationship between True Stress–True Strain Behavior and Deformation Microstructure in the Plastic Deformation of Neutron-Irradiated or Work-Hardened Austenitic Stainless Steel

True stress–true strain relation and deformation microstructure have been examined for high purity Fe-18Cr-12Ni alloy and its alloys doped with 0.7 wt % Si or 0.09 wt % C. In high purity alloy and C-doped alloy irradiated at 240°C up to 3 dpa, the work hardening rate is equivalent to that in unirradiated alloys. These alloys show dislocation channel structure after irradiation and deformation. In irradiated Si-doped alloy, however, the work hardening rate is different from that in unirradiated alloys. This alloy shows fully developed dislocation cell structure after deformation, as seen in unirradiated deformed stainless steels. The cell structure in irradiated Si-doped alloy was much smaller than that in unirradiated Si-doped alloy and in type 316L stainless steel. One of the factors affecting the change in the work hardening rate of irradiated austenitic stainless steel at 240°C is strong obstacles such as γ precipitate that acts as dislocation pining and dislocation loops such as Frank loops that do not act as obstacles.

Simulation for radiolytic products of seawater: effects of seawater constituents, dilution rate, and dose rate

Radiolysis calculations of simulated seawater were conducted using reported data on chemical yields and chemical reaction sets to predict the effects of seawater constituents on water radiolysis. Hydrogen, oxygen, and hydrogen peroxide were continuously produced from simulated seawater during γ-ray irradiation. The concentration of H2 exceeded its saturation concentration before it reached the steady-state concentration. The production behavior of these molecules was significantly promoted by the addition of bromide ions (Br−) because of the high reactivity of Br− with the hydroxyl radical, an effective hydrogen scavenger. It is also shown that the concentrations of these molecules were effectively suppressed by diluting seawater constituents by less than 1%.

Stress Corrosion Cracking Behavior of Type 304 Stainless Steel Irradiated under Different Neutron Dose Rates at JMTR

In order to investigate the effect of neutron dose rate on tensile properly and irradiation stress corrosion cracking (IASCC) behavior, crack growth rate (CGR) and, tensile tests and microstructure observation have been conducted with type 304 stainless steel specimens. The specimens were irradiated in high temperature water simulating boiling water reactor (BWR) environments up to about 1dpa with two different dose rates at the Japan Materials Testing Reactor (JMTR). While radiation hardening increased with the dose rate, CGR was not affected by the dose rate. Increase of the yield strength of the low dose rate specimens was caused by the increase of number density of Frank loops. Little difference of radiation-induced segregation at grain boundaries was observed in specimens irradiated by different dose rates. Furthermore, no dose rate effect on local plastic deformation behavior was found near crack tip in the crystal plasticity simulation

Technical Development for IASCC Irradiation Experiments at the JMTR

Irradiation assisted stress corrosion cracking (IASCC) is considered to be one of the key issues from a viewpoint of the life management of core components in the aged Light Water Reactors (LWRs). To simulate IASCC behavior by the in-pile IASCC experiment or post-irradiation experiment (PIE), it is necessary to irradiate specimens up to a neutron fluence that is higher than the so-called IASCC threshold fluence in a test reactor. There are, however, some technical hurdles to overcome for the experiments. For the in-pile IASCC test, techniques assembling pre-irradiated specimens into an in-pile test capsule in a hot cell by remote handling are necessary, and the Japan Atomic Energy Agency (JAEA) developed the techniques for the in-pile test to be carried out in the Japan Material Testing Reactor (JMTR). To examine crack growth and crack initiation behaviors under neutron irradiation, pre-irradiated specimens were relocated from pre-irradiation capsules to an in-pile capsule. Hence, a remote welding machine has been newly developed and welding work for inner and outer tubes of capsule are carried out with rotating of the capsule. The other hurdle is the material integrity of the capsule of the capsule housing for a long term irradiation. Since the changes in microstructure, micro chemistry and mechanical properties of materials increase with neutron fluence, the integrity for capsules of long irradiation period was evaluated by tensile tests in the air and slow strain rate test (SSRT) in oxygenated water. Specimens were obtained from the outer tubes of capsule irradiated to 1.0–3.9 × 1026 n/m2 (E> 1 MeV) previously. Elongation more than 15% in tensile test at 423 K was confirmed and no IGSCC fraction was shown in SSRT at 423 K which was estimated as temperature at the outer tubes of the capsule under irradiation.Copyright