Motomasa Fuse

Evaluation of self-interstitial properties in hcp zirconium using a computer simulation

Abstract The self-interstitial properties in hcp zirconium are investigated using dynamical-method computer simulations. The most stable self-interstitial is found to be located at the centroid of the open triangle in the basal plane, while the next most stable one is a crowdion type located at the point between two nearest-neighbor atoms in successive basal planes. Experimental results of annealing of radiation damage and He + back-scattering also suggest that the two types of configurations are the most possible candidates for self-interstitial in hcp zirconium. Based on the analysis on the configurations and the migration processes of self-interstitials, the recombination coefficients of vacancies and interstitials are evaluated.

Application of Hydrogen Water Chemistry to Moderate Corrosive Circumstances around the Reactor Pressure Vessel Bottom of Boiling Water Reactors

Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the reactor pressure vessel (RPV) and (b) moderating the crack growth rate, even if microcracks are present on the structural materials. Several disadvantages caused by HWC (e.g., turbine dose rate increase, 60 Co radioactivity buildup, and effects on fuel cladding) are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated based on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants - e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom - are planned that would collect data to support the theoretical considerations

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (I) Temperature Dependence of Hydrazine Reactions

Hydrazine and hydrogen co-injection into reactor water is considered a new mitigation method of stress corrosion cracking in BWRs. Fundamental data such as the thermal decomposition of hydrazine, the reaction of hydrazine with oxygen and with hydrogen peroxide at temperatures ranging from 150 to 280°C are needed to evaluate suitability of this method. Reactions in bulk water were studied in a polytetrafluoroethylene pipe to separate surface reaction effects. The results were as follows. (1) The orders of the apparent reaction rate of hydrazine with oxygen were 1 and 0.5 for hydrazine and oxygen concentrations, respectively . Arrhenius parameters were k 0=69.0 s−1.μM−0.5 and (2) The orders of apparent reaction rate of hydrazine with hydrogen peroxide were each 0.5 for hydrazine and hydrogen peroxide concentrations kC 0.5 N2H4 C 0.5 H2O2 . Arrhenius parameters were k 0=1.42x 106 s−1, Ea =78.8 kJ.mol−1. Based on these data, the applicability of hydrazine and hydrogen co-injection into BWRs was considered. Hydrazine introduction to reactor water was confirmed to be accompanied by only 1% decomposition. The concentration of oxygen, which is injected to suppress the flow-assisted corrosion of carbon steel in current BWR operation, would decrease due to the reaction of hydrazine with oxygen. However oxygen concentration in feed water could be maintained at the required level if the concentration of oxygen injected in condensate water was at most doubled compared to the current operating concentration.

Development of a Suppression Method for Deposition of Radioactive Cobalt after Chemical Decontamination: (I) Effect of the Ferrite Film Coating on Suppression of Cobalt Deposition

In the last decade, chemical decontamination at the beginning of periodical inspection has been applied to many Japanese BWR plants in order to reduce radiation exposure. However, following the chemical decontamination, a rapid dose rate increase can be seen in some plants after just a few operation cycles. Oxide film, which easily incorporates radioactivity, might be formed after the chemical decontamination. We developed a new way to reduce the recontamination after the chemical decontamination to maintain long-term continued decontamination effects without any chemical injections or chemical controls in reactor water during operation. In our approach, a fine ferrite film is formed by the Hitachi Ferrite Coat process after oxide films formed during the plant operation are removed by the chemical decontamination process.We select Fe(HCOO)2 aqueous solution, H2O2, and N2H4 as the treatment chemicals for fine ferrite film formation for suitable BWR plant application. Our laboratory experiment results confirm a 60Co deposition reduction effect of 1/5 compared with that of nontreatment for up to 3,100 hours. The fine ferritefilm that was formed on the specimen before the 60Co deposition test remains as a film structure after the test. The corrosion amount of the specimen is suppressed to 1/4 through the effect of the fine ferrite film.

Effects of alloying elements (Mo, Si) in an austenitic stainless steel on dislocation loop nucleation under ion irradiation

Abstract Effects of alloying elements in an austenitic stainless steel on dislocation loop formation under 300 keV He+ irradiation were studied. Pure stainless steel (Fe-18 wt.% Cr-16 wt.% Ni), Mo added and Si- and/or Mo-added pure stainless steels and Type 316 stainless steel and that without Mo were used. The numbers of dislocation loops at several temperatures between 250 and 450°C were measured. For pure stainless steel, activation energy of dislocation loop nucleation is obtained as 6 × 10−20 J (0.4 eV). Mo addition does not have any effects on loop nucleation, while Si addition to the pure stainless steel promotes loop nucleation and increases the activation energy to 8 × 10−20J (0.5 eV). But when Mo is added to Fe-Cr-Ni-Si alloy, loop density and activation energy decrease to those of the pure stainless steel. These experimental results lead to the conclusion that Mo addition suppresses heterogeneous loop nucleation, in which Si acts as the nucleus, by forming Mo-Si clusters, and these ideas are used to explain differences in mechanical properties under irradiation between Types 304 and 316 stainless steels.

Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (IV) : Reaction Mechanism and Plant Feasibility Analysis

A calculation model has been developed in order to evaluate effectiveness of hydrazine and hydrogen co-injection (HHC) into reactor water for mitigation of intergranular stress corrosion cracking of structural materials used in boiling water reactors (BWRs). The HHC uses the strong reducing power of hydrazine radical, which is produced in the downcomer region under irradiation by γ-rays and neutrons. Some reactions and their reaction rate constants were determined based on experiments which were carried out in aerated water, hydrogenated water, and deaerated water. The calculated results were in good agreement with experimental data by a factor of two. The model was applied to a BWR and it was found that the HHC cut oxygen and hydrogen peroxide amounts dissolved in reactor water more effectively than hydrogen water chemistry alone. Thus, the required amount of hydrogen for hydrazine injection was much lower than that for hydrogen water chemistry. Consequently, electrochemical corrosion potential of structural materials could be lowered below–0:1V vs. SHE without any increase of MS line dose rate, which has been a limitation of the conventional hydrogen water chemistry. The HHC was predicted to decrease crack growth rate of structural materials by a factor of 10.

Study of Polarization Curve Measurement Method for Type 304 Stainless Steel in BWR High Temperature-High Purity Water

In order to improve the accuracy of electrochemical corrosion potential (ECP) calculations, information about the effects of the flow velocity, the type of oxidization species, the oxide film thickness, and the oxide film classification on the polarization curve is needed. Polarization curve measurements were studied here with that information goal in mind. A wire-shaped working electrode was employed to remove the IR drop caused by high solution resistance, which makes it difficult to apply a proper potential to the working electrode. In order to measure the static state current, the potential scanning condition was optimized.It was confirmed that a steady-state current can be measured by establishing a potential scanning rate for applying a potential not greater than 0.01mV.s−1 with a step-shaped waveform. This method was employed to measure the anodic polarization curve of type 304 stainless steel in deaerated 553K high-purity water. The passivation behavior was clearly observed. Also, the cathodic polarization curves in water containing dissolved oxygen of less than 1000 ppb were measured. The oxygen concentration dependence of ECP was calculated from measured polarization curves. The ECPs calculated using the measured polarization curves were in good agreement with those measured with an electrometer, confirming the validity of the measured polarization curves.

Hydrazine and hydrogen Co-injection to mitigate stress corrosion cracking of structural materials in boiling water reactors, (II) : Reactivity of hydrazine with oxidant in high temperature water under gamma-irradiation

Hydrogen and hydrazine co-injection into a boiling water reactor was considered as a new mitigation method of stress corrosion cracking. To confirm decrease of electrochemical corrosion potential by the reduction of oxygen and hydrogen peroxide in bulk water using reducing agent, reaction of hydrazine with oxygen or hydrogen peroxide under simulated downcomer conditions (temperature: 280°C, duration: 4.2 s and gamma-irradiation) was examined. All the oxygen was consumed above the equivalent concentration for the reaction with oxygen within this short time (gamma-irradiation case) and almost all the hydrogen peroxide was consumed. Reaction rates were accelerated more than five times by gamma-irradiation in each case. Reaction rates with oxygen were compared with other reducing agents such as hydrogen, methanol and ammonia. From the viewpoint of reaction rate and formation of by-product, hydrazine was the most suitable agent. Using a simple model based on the experimental results, water chemistry for the bottom region was calculated for the case of no hydrogen injection and for the case of 0.2 mmol-kg−1 hydrogen injection into feed water. For both cases, concentrations of dissolved oxygen and hydrogen peroxide were estimated to decrease enough to mitigate SCC, with concentration of ammonia suppressed below the management criteria for reactor water chemistry.

Analysis of irradiation growth in zirconium-base alloy

Abstract Irradiation induced growth phenomena in zirconium-base alloy have been investigated on the basis of the numerical results obtained by dynamical-method computer simulations. Point defect trapping by solute atoms causes transient growth strain which is recoverable during temperature cycling. The growth strain change of cold-worked Zircaloy-2 irradiated at 553 K can be explained by a dislocation climb and a net inflow of vacancies into grain boundaries. The effects of various parameters such as grain size are also discussed.

Hydrazine and Hydrogen Coinjection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors (VII)—Effects of Bulk Water Chemistry on ECP Distribution inside a Crack

Water chemistry in a simulated crack (crack) has been studied to understand the mechanisms of stress corrosion cracking in a boiling water reactor environment. Electrochemical corrosion potential (ECP) in a crack made in an austenite type 304 stainless steel specimen was measured. The ECP distribution along the simulated crack was strongly affected by bulk water chemistry and bulk flow. When oxygen concentration was high in the bulk water, the potential difference between the crack tip and the outside of the crack (ΔE), which must be one motive force for crack growth, was about 0.3V under a stagnant condition. When oxygen was removed from the bulk water, ECP inside and outside the crack became low and uniform and ΔE became small. The outside ECP was also lowered by depositing platinum on the steel specimen surface and adding stoichiometrically excess hydrogen to oxygen to lower ΔE. This was effective only when bulk water did not flow. Under the bulk water flow condition, water-borne oxygen caused an increase in ECP on the untreated surface inside the crack. This also caused a large ΔE. The ΔE effect was confirmed by crack growth rate measurements with a catalyst-treated specimen. Therefore, lowering the bulk oxidant concentration by such measures as hydrazine hydrogen coinjection, which is currently under development, is important for suppressing the crack growth.