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Stefan Ritter - One of the best experts on this subject based on the ideXlab platform.
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the influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low alloy steels under simulated Boiling Water Reactor conditions
Corrosion Science, 2016Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy Reactor pressure vessel steels was evaluated under simulated Boiling Water Reactor conditions. In normal Water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity Water values within few 100 h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen Water chemistry environment is much higher.
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the influence of ppb levels of chloride impurities on the strain induced corrosion cracking and corrosion fatigue crack growth behavior of low alloy steels under simulated Boiling Water Reactor conditions
Corrosion Science, 2016Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The effect of chloride on the strain-induced corrosion cracking (SICC) and corrosion fatigue (CF) crack growth behaviour in low-alloy Reactor pressure vessel steels was evaluated under simulated Boiling Water Reactor normal Water chemistry conditions by slow rising load and cyclic constant load amplitude tests with air fatigue pre-cracked fracture mechanics specimens. Chloride in the ppb level range increased the SICC initiation susceptibility, but had almost no effect on the subsequent SICC and CF crack growth. A strong effect of chloride addition of 100 ppb on CF crack growth was observed at intermediate corrosion potentials and very low loading frequencies only.
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strain induced corrosion cracking behaviour of low alloy steels under Boiling Water Reactor conditions
Journal of Nuclear Materials, 2008Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The strain-induced corrosion cracking (SICC) behaviour of different low-alloy Reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated Boiling Water Reactor (BWR)/normal Water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP ⩾+50 mV SHE , ⩾0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness K I value of 25 MPa m 1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate d K I /d t , but were not dependent on the actual K I values up to 60 MPa m 1/2 . A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200–250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.
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corrosion fatigue crack growth behaviour of low alloy Reactor pressure vessel steels under Boiling Water Reactor conditions
Corrosion Science, 2008Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The corrosion fatigue crack growth behaviour of different low-alloy Reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was systematically characterized under simulated Boiling Water Reactor normal Water and hydrogen Water chemistry conditions by low-frequency fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing Water at temperatures from 150 to 288 °C. In this paper, the observed synergistic effects of environmental, material and loading parameters on the environmental acceleration of fatigue crack growth in low-alloy RPV steels are discussed in the context of the Ford–Andresen model. Additionally, the adequacy and conservatism of the current “ASME XI reference fatigue crack growth curves” of the ASME Boiler & Pressure Vessel Code are critically reviewed and assessed on the basis of the gathered experimental data base and this model. Based on the observed cracking behaviour and the Ford–Andresen model, a simple time-domain superposition model is suggested, which could reduce most of the undue conservatism and eliminate uncertainties of the existing codes and therefore serve as a basis for the development of improved reference fatigue crack growth curves.
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stress corrosion cracking of low alloy Reactor pressure vessel steels under Boiling Water Reactor conditions
Journal of Nuclear Materials, 2008Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The stress corrosion cracking (SCC) behaviour of different Reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated Boiling Water Reactor (BWR) normal Water (NWC) and hydrogen Water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing Water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (
H P Seifert - One of the best experts on this subject based on the ideXlab platform.
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the influence of ppb levels of chloride impurities on the stress corrosion crack growth behaviour of low alloy steels under simulated Boiling Water Reactor conditions
Corrosion Science, 2016Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The effect of chloride on the stress corrosion crack (SCC) growth behaviour in low-alloy Reactor pressure vessel steels was evaluated under simulated Boiling Water Reactor conditions. In normal Water chemistry environment, ppb-levels of chloride may result in fast SCC after rather short incubation periods of few hours. After moderate and short-term chloride transients, the SCC crack growth rates return to the same very low high-purity Water values within few 100 h. Potential long-term (memory) effects on SCC crack growth cannot be excluded after severe and prolonged chloride transients. The chloride tolerance for SCC in hydrogen Water chemistry environment is much higher.
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the influence of ppb levels of chloride impurities on the strain induced corrosion cracking and corrosion fatigue crack growth behavior of low alloy steels under simulated Boiling Water Reactor conditions
Corrosion Science, 2016Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The effect of chloride on the strain-induced corrosion cracking (SICC) and corrosion fatigue (CF) crack growth behaviour in low-alloy Reactor pressure vessel steels was evaluated under simulated Boiling Water Reactor normal Water chemistry conditions by slow rising load and cyclic constant load amplitude tests with air fatigue pre-cracked fracture mechanics specimens. Chloride in the ppb level range increased the SICC initiation susceptibility, but had almost no effect on the subsequent SICC and CF crack growth. A strong effect of chloride addition of 100 ppb on CF crack growth was observed at intermediate corrosion potentials and very low loading frequencies only.
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strain induced corrosion cracking behaviour of low alloy steels under Boiling Water Reactor conditions
Journal of Nuclear Materials, 2008Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The strain-induced corrosion cracking (SICC) behaviour of different low-alloy Reactor pressure vessel (RPV) and piping steels and of a RPV weld filler/weld heat-affected zone (HAZ) material was characterized under simulated Boiling Water Reactor (BWR)/normal Water chemistry (NWC) conditions by slow rising load (SRL) and very low-frequency fatigue tests with pre-cracked fracture mechanics specimens. Under highly oxidizing BWR/NWC conditions (ECP ⩾+50 mV SHE , ⩾0.4 ppm dissolved oxygen), the SICC crack growth rates were comparable for all materials (hardness K I value of 25 MPa m 1/2 had to be exceeded to initiate SICC in SRL tests. Above this value, the SICC rates increased with increasing loading rate d K I /d t , but were not dependent on the actual K I values up to 60 MPa m 1/2 . A maximum in SICC initiation susceptibility occurred at intermediate temperatures around 200–250 °C and at slow strain rates in all materials. In contrast to crack growth, the SICC initiation susceptibility was affected by environmental and material parameters within certain limits.
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corrosion fatigue crack growth behaviour of low alloy Reactor pressure vessel steels under Boiling Water Reactor conditions
Corrosion Science, 2008Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The corrosion fatigue crack growth behaviour of different low-alloy Reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was systematically characterized under simulated Boiling Water Reactor normal Water and hydrogen Water chemistry conditions by low-frequency fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing Water at temperatures from 150 to 288 °C. In this paper, the observed synergistic effects of environmental, material and loading parameters on the environmental acceleration of fatigue crack growth in low-alloy RPV steels are discussed in the context of the Ford–Andresen model. Additionally, the adequacy and conservatism of the current “ASME XI reference fatigue crack growth curves” of the ASME Boiler & Pressure Vessel Code are critically reviewed and assessed on the basis of the gathered experimental data base and this model. Based on the observed cracking behaviour and the Ford–Andresen model, a simple time-domain superposition model is suggested, which could reduce most of the undue conservatism and eliminate uncertainties of the existing codes and therefore serve as a basis for the development of improved reference fatigue crack growth curves.
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stress corrosion cracking of low alloy Reactor pressure vessel steels under Boiling Water Reactor conditions
Journal of Nuclear Materials, 2008Co-Authors: H P Seifert, Stefan RitterAbstract:Abstract The stress corrosion cracking (SCC) behaviour of different Reactor pressure vessel (RPV) steels and weld filler/heat-affected zone materials was characterized under simulated Boiling Water Reactor (BWR) normal Water (NWC) and hydrogen Water chemistry (HWC) conditions by periodical partial unloading, constant and ripple load tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated or hydrogenated high-purity or sulphate/chloride containing Water at temperatures from 150 to 288 °C. In good agreement with field experience, these investigations revealed a very low susceptibility to SCC crack growth and small crack growth rates (
Tetsuo Shoji - One of the best experts on this subject based on the ideXlab platform.
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quantitative prediction of eac crack growth rate of sensitized type 304 stainless steel in Boiling Water Reactor environments based on epfem
Journal of Pressure Vessel Technology-transactions of The Asme, 2007Co-Authors: Tetsuo ShojiAbstract:The quantitative prediction of environmentally assisted cracking (EAC) or stress corrosion cracking (SCC) is essential in order to predict service life and also the structural integrity and safety assessment of light Water Reactors. During the last 3 decades many of the research results obtained on the quantitative prediction of the EAC crack growth rate have been based on linear fracture mechanics. In order to investigate EAC behavior in the high strain zone of important structures in light Water Reactors, the approach taken in this paper is one in which quantitative calculations of the EAC crack growth rate, incorporating the SCC deformation /oxidation model and the elastic-plastic finite element method (EPFEM), are carried out. This approach can be used for the quantitative prediction of EAC crack growth rate in both the low and high strain zones of key structures in light Water Reactors. The crack growth behavior of sensitized type 304 stainless steel with a 1T-CT specimen in simulated Boiling Water Reactor (BWR) environments is analyzed based on this approach. The effect of several environmental, material, and mechanical parameters on the EAC crack growth rate of nickel based alloys in high-temperature aqueous environments is also discussed.
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intergranular environmentally assisted cracking of alloy 182 weld metal in simulated normal Water chemistry of Boiling Water Reactor
Corrosion Science, 2007Co-Authors: Qunjia Peng, Tetsuo Shoji, Hiroyuki Yamauchi, Yoichi TakedaAbstract:Abstract Following characterization of grain boundaries in Alloy 182 weld metal, intergranular environmentally assisted cracking (IGEAC) growth behavior of the alloy in simulated normal Water chemistry (NWC) of Boiling Water Reactor (BWR) was tested by employing a crack growth rate test. The grain boundary was found consist of about 72% random boundaries, 21% low-angle boundaries and a few coincident boundaries. The correlation between the grain boundary characters and IGEAC was discussed. Based on the IGEAC growth rates obtained in the test and evaluations of the crack tip strain rate, the parameter of the slip-dissolution/oxidation model that embodies the kinetics of oxidation/repassivation in the crack-tip environment of Alloy 182 in BWR-NWC was quantitatively determined.
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The Crack Tip Solution Chemistry in Sensitized Stainless Steel in Simulated Boiling Water Reactor Water Studied Using a Microsampling Technique
Journal of Nuclear Science and Technology, 2003Co-Authors: Qunjia Peng, Tetsuo ShojiAbstract:In order to obtain a better understanding of the mechanism of stress corrosion cracking, an improved microsampling technique was applied to measurements of the local solution chemistry at the growing tip of a stress corrosion crack. Efforts were made to improve the microsampling technique to obtain more accurate information about the crack tip solution chemistry and apply it to measurements of the crack tip solution chemistry in sensitized 304L stainless steel in simulated Boiling Water Reactor Water. The improved technique was used to study the effects of dissolved oxygen, testing temperature and exposure time on the concentration of harmful anions at the crack tip. The microsampled solution from the crack tip was analyzed by a Capillary Electrophoresis system. The results as well as the effects of the improved technique were discussed.
Mamoru Ishii - One of the best experts on this subject based on the ideXlab platform.
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a core design study for a small modular Boiling Water Reactor with long life core
Nuclear Technology, 2016Co-Authors: W S Yang, Shanbin Shi, Mamoru IshiiAbstract:This paper presents the core design and performance characteristics of the Novel Modular Reactor (NMR-50), a 50-MW(electric) small modular Reactor. NMR-50 is a Boiling Water Reactor with na...
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Design and test of hydraulic vacuum breaker check valve for simplified Boiling Water Reactor
Nuclear Engineering and Design, 2006Co-Authors: H.j. Yoon, Shripad T. Revankar, Mamoru IshiiAbstract:In the current design of the simplified Boiling Water Reactor, the vacuum breaker check valve is an important safety component. The vacuum breaker check valve is the only key safety components which is not passive in nature. Failure of this mechanical valve drastically reduces the passive containment cooling system cooling capability and hence containment pressure may exceed the design pressure. To eliminate this problem novel vacuum breaker check valve was developed to replace the mechanical valve. This new design is based on a passive hydraulic head, which is fail-safe and is truly passive in operation. Moreover this new design needs only one additional tank and one set of piping each to the wetwell and drywell. This system is simple in design and hence is easy to maintain and to qualify for operation. The passive vacuum breaker check valve performance was first evaluated using RELAP5. Then the passive vacuum breaker check valve was constructed and implemented in the PUMA integral test facility. Its performance was studied in a large break loss of coolant accident simulation test performed in PUMA facility.
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Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor
10th International Conference on Nuclear Engineering Volume 2, 2002Co-Authors: Mamoru Ishii, Shripad T. RevankarAbstract:A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified Boiling Water Reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified Boiling Water Reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident.Copyright © 2002 by ASME
Wenjun Kuang - One of the best experts on this subject based on the ideXlab platform.
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the effect of post irradiation annealing on the stress corrosion crack growth rate of neutron irradiated 304l stainless steel in Boiling Water Reactor environment
Corrosion Science, 2019Co-Authors: Wenjun Kuang, Justin HesterbergAbstract:Abstract The effect of post-irradiation annealing (PIA) on the irradiation-assisted stress corrosion cracking (IASCC) of 304L stainless steel in Boiling Water Reactor environments was studied. PIA has little effect on the crack growth rate (CGR) in normal Water chemistry even it restored most of the irradiation-induced microstructure change. Nevertheless, PIA can significantly suppress the CGR in hydrogen Water chemistry (HWC). The mitigation factor of HWC increases with the degree of PIA up to 5 h at 550 °C and correlates with the recovery of Si segregation at grain boundary. PIA is an effective strategy for decreasing IASCC susceptibility in reducing Water environment.