• Corrosion Science and Technology
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• 제7권2호
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• pp.77-84
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• 2008

Based upon the good compatibility to neutron irradiation and high temperature water environment, austenitic stainless steels are widely used for core internal structural materials of light water reactors. But, recently, intergranular cracking was detected in the stainless steels for the core applications in some commercial PWR plants. Authors studied on the root cause of the intergranular cracking and developed the countermeasure including the alternative materials for these core applications. The intergranular cracking in these core applications are defined as an irradiation assisted mechanical cracking and irradiation assisted stress corrosion cracking. In this paper, the root cause of the intergranular cracking and its countermeasure are summarized and discussed.

• 대한금속재료학회지
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• 제50권8호
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• pp.583-589
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• 2012

This work is concerned with a statistical approach to the life prediction on irradiation-assisted stress corrosion cracking (IASCC) of stainless steel (SS) for core internals of a pressurized water reactor (PWR). The previous results of the time-to-failure of IASCC measured on neutron-irradiated stainless steel components were statistically analyzed in terms of stress and irradiation. The accelerating life testing model of IASCC of cold worked Type 316 SS was established based on an inverse power model with two stress-variables, the applied stress and irradiation dose. Considering the variation of the yield strength and applied stress with the irradiation dose in the model, the remaining life of the baffle former bolt was statistically predicted during operation under complex environments of stress and irradiation.

• 대한금속재료학회지
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• 제47권12호
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• pp.819-827
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• 2009

This work is concerned with a statistical analysis of factors affecting the irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels for core internals of pressurized water reactors (PWR). The microstructural and environmental factors were reviewed and critically evaluated by the statistical analysis. The Cr depletion at grain boundary was determined to have no significant correlation with the IASCC susceptibility. The threshold irradiation fluence of IASCC in a PWR was statistically calculated to decrease from 5.799 to 1.914 DPA with increase of temperature from 320 to $340^{\circ}C$. From the analysis of the relationship between applied stress and time-to-failure of stainless steel components based on an accelerated life testing model, it was found that B2 life of a baffle former bolt exposed to neutron fluence of 20 and 75 DPA was at least 2.5 and 0.4 year, respectively, within 95% confidence interval.

• Corrosion Science and Technology
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• 제20권4호
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• pp.210-229
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• 2021

To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.

• Corrosion Science and Technology
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• 제20권3호
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• pp.158-168
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• 2021

Austenitic 316 stainless steel was irradiated with protons accelerated by an energy of 2 MeV at 360 ℃, the various defects induced by this proton irradiation were characterized with microscopic equipment. In our observations irradiation defects such as dislocations and micro-voids were clearly revealed. The typical irradiation defects observed differed according to depth, indicating the evolution of irradiation defects follows the characteristics of radiation damage profiles that depend on depth. Surface oxidation tests were conducted under the simulated primary water conditions of a pressurized water reactor (PWR) to understand the role irradiation defects play in surface oxidation behavior and also to investigate the resultant irradiation assisted stress corrosion cracking (IASCC) susceptibility that occurs after exposure to PWR primary water. We found that Cr and Fe became depleted while Ni was enriched at the grain boundary beneath the surface oxidation layer both in the non-irradiated and proton-irradiated specimens. However, the degree of Cr/Fe depletion and Ni enrichment was much higher in the proton-irradiated sample than in the non-irradiated one owing to radiation-induced segregation and the irradiation defects. The microstructural and microchemical changes induced by proton irradiation all appear to significantly increase the susceptibility of austenitic 316 stainless steel to IASCC.

• Corrosion Science and Technology
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• 제22권6호
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• pp.457-465
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• 2023

Irradiation-assisted stress corrosion cracking (IASCC) is one of the main degradation mechanisms of austenitic stainless steels, which are used as reactor internal materials. Slow strain rate testing (SSRT) has been widely applied to evaluate the IASCC initiation characteristics of proton-irradiated tensile specimens. Tensile specimens require low surface roughness for micro-crack observation, and electropolishing is the most important specimen pre-treatment process used for this. In this study, optimal electropolishing conditions were examined through analyzing results of polarization experiments and surface roughness measurements after electropolishing. Corrosion cell and electropolishing equipment were fabricated for polarization tests and electropolishing experiments using SSRT specimens. The experimental parameters were electropolishing time, current density, electrolyte temperature, and stirring speed. The optimal electropolishing conditions for SSRT tensile specimens made of type 316 stainless steel were evaluated as a polishing time of 180 seconds, a current density of 0.15 A/cm², an electrolyte temperature of 60 ℃, and a stirring speed of 200 RPM.

• Nuclear Engineering and Technology
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• 제41권9호
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• pp.1171-1180
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• 2009

The internal core baffle structure of a French Pressurized Water Reactor (PWR) consists of a collection of baffles and formers that are attached to the barrel. The connections are done thanks to a large number of bolts (about 1500). After inspection, some of the bolts have been found cracked. This has been attributed to the Irradiation Assisted Stress Corrosion Cracking (IASCC). The $Electricit\acute{e}$ De France (EDF) has set up a research program to gain better knowledge of the temperature distribution, which may affect the bolts and the whole structure. The temperature distribution in the structure was calculated thanks to the thermal code SYRTHES that used a finite element approach. The heat transfer between the by-pass flow inside the cavities of the core baffle and the structure was accounted for thanks to a strong thermal coupling between the thermal code SYRTHES and the CFD code named Code_Saturne. The results for the CP0 plant design show that both the high temperature and strong temperature gradients could potentially induce mechanical stresses. The CPY design, where each bolt is individually cooled, had led to a reduction of temperatures inside the structures. A new parallel version of SYRTHES, for calculations on very large meshes and based on MPI, has been developed. A demonstration test on the complete structure that has led to about 1.1 billion linear tetraedra has been calculated on 2048 processors of the EDF Blue Gene computer.