Boiling Water Reactor Components Research Articles

Oxide Film Characterization after the crack propagation in CT specimens of AISI 304L under Hydrogen Water Chemistry Condition.

ABSTRACTStress Corrosion Cracking (SCC) in a general term describing stressed alloy fracture that occurs by crack propagation in specifically environments, and has the appearance of brittle fracture, yet it can occur in ductile materials like AISI 304L used in internal components of Boiling Water Reactors (BWR). The high levels of oxygen and hydrogen peroxide generated during an operational Normal Water Condition (NWC) promotes an Electrochemical Corrosion Potential (ECP), enough to generate SCC in susceptible materials. Changes in water chemistry have been some of the main solutions for mitigate this degradation mechanism, and one of these changes is reducing the ECP by the injection of Hydrogen in the feed water of the reactor; this addition moves the ECP below a threshold value, under which the SCC is mitigated (-230mV vs SHE). This paper shows the characterization by Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD) and Raman Spectroscopy of the oxide film formed in to a crack propagated during a Rising Displacement Test method (RDT), on Hydrogen Water Chemistry (HWC) conditions: 20 ppb O2, 125 ppb H2, P=8MPa, T=288°C, using a CT specimen of austenitic stainless steel AISI 304L sensitized. The characterization allowed identifying the magnetite formation since an incipient way, until very good formed magnetite crystals.

High Pressure Boiling Water Reactor

Some 400 boiling water reactors (BWRs) and pressurized water reactors (PWRs) have been in operation for several decades. The presented concept, the high pressure boiling water reactor (HP-BWR), combines the best parts and omits the troublesome components of traditional BWRs and PWRs by taking into consideration the experiences gained during their operation.One of the major benefits of the HP-BWR is that safety is improved. The design utilizes gravity-operated control rods, and there is a large space for the cross-formed control rods between fuel boxes. The bottom of the reactor vessel is smooth and without penetrations. All the pipe connections to the reactor vessel are well above the top of the reactor core, and core spray is not needed. Additionally, internal circulation pumps are used.The HP-BWR concept is also environmentally friendly: Improved thermal efficiency is achieved by feeding the turbine with ~340°C (15 MPa) steam instead of ~285°C (7 MPa), and there is less warm water release to the recipient and less uranium consumption per produced kWh, resulting in the production of less waste.Finally, the HP-BWR is cost effective and simple, operating in direct cycle mode with no need for complicated steam generators. Moisture separators and steam dryers are placed inside the reactor vessel, and additional separators and dryers can be installed inside or outside the containment. Well-proved simple dry containment or wet containment can be used.In more than half a century, an extensive regulatory licensing experience has been built from traditional BWRs and PWRs. The HP-BWR is a developed, high-performance successor of those conventional designs. Therefore, it can be expected that licensing can be accomplished in a reasonable time.Several utilities are supporting manufacturers to study concepts for future reactors. It is likely that an application to one or more electrical power companies for financial support by a manufacturer to make a detailed feasibility study of the HP-BWR would be positively treated. This could be the next step to the implementation of the HP-BWR.

The Influence of Various Treating Parameters on the Corrosion of ZrO2-Treated Type 304 Stainless Steels in High-Temperature Water

Corrosion mitigation techniques specifically developed for the vessel internals of boiling water reactors (BWRs) such as hydrogen water chemistry (HWC), catalytic coatings, and inhibitive coatings have been studied and/or adopted worldwide in the past decade. The common goal of these techniques is to mitigate intergranular stress corrosion cracking in the vessel internal components of BWRs. For enhancing the effectiveness of HWC through a more effective reduction in electrochemical corrosion potential accompanied by a lower hydrogen consumption rate or even without any added hydrogen, the technique of inhibitive treatment on structural component surfaces was brought into consideration. In the current study, surfaces of preoxidized square-disk specimens made of Type 304 stainless steel (SS) were treated with zirconium oxide (ZrO2) (particle sizes of 0.7 μm and 0.1 μm) via hydrothermal deposition at either 90 or 150°C for 7 or 14 d, while some preoxidized specimens remained untreated. The concentration of the treating solution was maintained at either 10 parts per million (ppm) or 122 ppm (1mM) during the ZrO2 deposition processes. Electrochemical polarization analyses at 288°C were conducted to characterize the electrochemical properties of the treated and untreated SS specimens in pure water with either dissolved oxygen or dissolved hydrogen. It was found that a smaller particle size, a longer treating time, a lower treating temperature, as well as a higher treating solution concentration seemed to improve the corrosion resistance of the treated specimens in oxygenated environments and to moderately deter the hydrogen oxidation rate on the specimens.

The Influence of Various Treating Parameters on the Corrosion of ZrO2-Treated Type 304 Stainless Steels in High-Temperature Water

Corrosion mitigation techniques specifically developed for the vessel internals of boiling water reactors (BWRs) such as hydrogen water chemistry (HWC), catalytic coatings, and inhibitive coatings have been studied and/or adopted worldwide in the past decade. The common goal of these techniques is to mitigate intergranular stress corrosion cracking in the vessel internal components of BWRs. For enhancing the effectiveness of HWC through a more effective reduction in electrochemical corrosion potential accompanied by a lower hydrogen consumption rate or even without any added hydrogen, the technique of inhibitive treatment on structural component surfaces was brought into consideration. In the current study, surfaces of preoxidized square-disk specimens made of Type 304 stainless steel (SS) were treated with zirconium oxide (ZrO2) (particle sizes of 0.7 μm and 0.1 μm) via hydrothermal deposition at either 90 or 150°C for 7 or 14 d, while some preoxidized specimens remained untreated. The concentration of the treating solution was maintained at either 10 parts per million (ppm) or 122 ppm (1mM) during the ZrO2 deposition processes. Electrochemical polarization analyses at 288°C were conducted to characterize the electrochemical properties of the treated and untreated SS specimens in pure water with either dissolved oxygen or dissolved hydrogen. It was found that a smaller particle size, a longer treating time, a lower treating temperature, as well as a higher treating solution concentration seemed to improve the corrosion resistance of the treated specimens in oxygenated environments and to moderately deter the hydrogen oxidation rate on the specimens.

Susceptibility of Sensitized Type 304 Stainless Steel to Intergranular Stress Corrosion Cracking in Simulated Boiling-Water Reactor Environments

Abstract Slow strain rate tests (SSRT) were conducted to evaluate the mitigating effect of hydrogen water chemistry (HWC) on intergranular stress corrosion cracking (IGSCC) of sensitized type 304 (UNS S30400) stainless steel (SS) used for in-reactor components of boiling-water reactors (BWR). The in-reactor environment was simulated by controlling water chemistry parameters such as dissolved oxygen (DO), dissolved hydrogen (DH), hydrogen peroxide (H2O2), and corrosion potential (Ecorr). Effects of the products of water radiolysis on SCC behavior were examined in the presence of a γ-radiation field. IGSCC susceptibility of sensitized type 304 SS decreased with decreasing concentrations of DO and H2O2, which suggested IGSCC of sensitized type 304 SS could be mitigated even in the in-reactor environment under HWC conditions. IGSCC susceptibility clearly was dependent upon Ecorr, as controlled by the concentrations of stable radiolytic products (e.g., O2, H2O2, and H2), regardless of short-lived radicals gene...

Core design for the boiling water reactor SWR 1000 / Nukleare Kernauslegung für den Siedewasserreaktor SWR 1000.

The SWR 1000 concept, which based on boiling water reactor components that have been proven in service, combines inherently passive safety with aspects of economical power generation, also affects the core design. The aim is to provide as simple a core configuration as possible and control rods and 648 fuel asssemblies of the ATRIUM 12 type. Detailed core analyses have demonstrated the feasibility of this concept

Viability of Hydrogen Water Chemistry for Protecting In-Vessel Components of Boiling Water Reactors

Abstract A mixed potential model (MPM) has been developed for calculating electrochemical potential (ECP) data for boiling water reactor (BWR) in-vessel components and recirculation piping using compositional data obtained from radiolysis water chemistry models. The calculated core component ECPs for the Duane-Arnold BWR are in good agreement with limited in-plant data. The calculated ECP data suggest that hydrogen water chemistry (HWC) will be effective in protecting the core inlet and the recirculation system of Duane-Arnold against intergranular stress corrosion cracking (IGSCC) but that it may not be effective for this purpose in Dresden-2 because of the higher concentrations of radiolysis products (notable O2 and H2O2) predicted to exist in this reactor. Furthermore, the MPM predicts that HWC will be ineffective in protecting various in-vessel components such as the upper plenum, fuel channels, core bypass structure, and the downcomer. Electrocatalysis of the redox reactions shows promise as a means ...

Fracture Mechanics Data and Modeling of Environmental Cracking of Nickel-Base Alloys in High-Temperature Water

Abstract Environmental cracking of ductile nickel-base alloys has occurred both in pressurized water reactors (e.g., steam-generator tubing) and boiling water reactor components such as pressure-vessel safe ends, weld butters, and filler metals for joining nickel-base alloys or dissimilar metals, and attachment welding pads on pressure vessels. Accurate assessment of the interrelated effects of material, environment, and mechanics on environmentally assisted crack growth rates is required for life prediction of nuclear power reactor components. The objective of this study is to understand and predict the environmental cracking behavior of ductile nickel-base alloys in 288° C water. Experiments were performed on 1T CT specimens of Alloy 600 base metal, and Alloy 182 and Alloy 82 weld metals to evaluate the effects of water chemistry, heat treatment, and stress intensity on crack growth rate. Fracture mechanics, crack growth data on ductile nickel-base alloys were also thoroughly surveyed to provide an over...

Proactive approaches to assure the structural integrity of Boiling Water Reactor components

Abstract Pressure vessel components in operating Boiling Water Reactor (BWR) plants are subjected to a variety of loading and environmental conditions which could lead to degradation over time. The significant damage mechanisms such as fatigue, stress corrosion cracking (SCC) and irradiation embrittlement are considered in the design basis of the reactor components and thus provide adequate structural margins over the operating life of the plant. Nevertheless, when the design basis assumptions are exceeded, e.g., thermal cycles, vibratory loading or chemistry transients, cracking may occur in pressure boundary components. Several proactive measures are being implemented to address this concern and assure the structural margins in BWR plants. These measures include: (i) control of materials and design to mitigate SCC and improvement of the environmental conditions through the implementation of Hydrogen Water Chemistry, (ii) advances in automated ultrasonic inspection of the BWR pressure vessel and piping, (iii) improved monitoring techniques for tracking fatigue usage and SCC effects in the piping and in the core, and (iv) development and qualification of durable repairs and specialized techniques such as use of high purity materials and temper bead repair. This paper describes current progress in implementing these proactive approaches for Boiling Water Reactors.

Oxidation and Radiation Buildup on Stainless Steel Components of Boiling Water Reactors

The buildup of radioactivity was measured on Type 304 stainless steel and carbon steel coupons exposed to reactor water in a test loop, which operated for ∼3 yr in a commercial boiling water reactor. Most of the activity buildup could be ascribed to 60Co. After an initial rapid increase, the activity buildup slowed down and kept increasing linearly up to 10 000 h of effective exposure to reactor water. No correlation was found between the 60Co activity in reactor water and the 60Co activity on the coupons. Preoxidation of Type 304 stainless steel coupons reduced the initial activity buildup. Carbon steel picked up much less activity than stainless steel during the tests. Measurements of the activity buildup after filtration of the primary coolant indicated that soluble impurities rather than particulates are primarily responsible for the radiation buildup phenomena. Increasing the coolant velocity also resulted in a reduction of the initial activity buildup on the coupons.