[논문]Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Water Reactors

Yonezawa, Toshio

Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Water Reactors 원문보기

Corrosion science and technology, v.7 no.2, 2008년, pp.77 - 84

Yonezawa, Toshio (Fracture and Reliability Research Institute, Tohoku University)

Abstract ▼ AI-Helper

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.

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문제 정의

In this paper, the results of the recent studies on IAMC and IASCC for the austenitic stainless steels in PWR environment are reviewed and the future direction of studies on this matter is discussed.

가설 설정

(d) PWSCC: Austenitic stainless steels have no susceptibility to IGSCC such as PWR primary water stress corrosion cracking (PWSCC) in Ni based alloys, as well known. However, the possibility of PWSCC for the irradiated austenitic stainless steel was indicated based upon the grain boundary chemical composition change by radiation induced segre- gation.
1) The effect of deformation of the absorber due to slumping caused by dynamic loads and creep is negligible.

제안 방법

for 2, 3 and 4 loop PWR plants. From our SCC tests under condensed alkaline caused by dry and wet phenomenon using by the model BFBs made of unirradiated 3 16 CW stainless steel, loaded actual stress and added with crevice devised under heat transfer condition at 331 to 332 C in simulated PWR primary water, IGSCC was observed at the shank area of the model BFBs, after 2, 000 hours test. From this result, if SCC due to condensed alkaline caused by dry and wet phenomenon will be able to occur in the actual BFBs, IGSCC will occur after the short time operation.
11) This defect does not directly harm the safety or operatability of the plant, but does give a large impact to world-wide utilities and fabricators of PWR plants. In order to establish the countermeasure for this cracking in the BFBs for PWR plants, author and previous co-workers studied on the cause of this IGC, and developed the alternative materials.
In order to evaluate the cause of the IGC in BFBs, above five possibilities were evaluated by the calculation and some out of pile tests from the metallurgical and chemical viewpoint.

성능/효과

2) The mechanism of the intergranular cracks in the cladding tube is not IASCC but IAMC caused by an increase in hoop strain due to swelling of the absorber at a very slow strain rate and decrease in elongation due to neutron irradiation.
This tendency looks-like the PWSCC susceptibility of Ni based alloys in simulated PWR primary water. It is concluded that the IGC of the highly irradiated austenitic stainless steel such as BFBs is PWSCC due to RIS.
5. The conclusion reached on the basis of the results of BWR measurements is that the hydrogen content is independent of both fast neutron fluence and flux. However, hydrogen content seems to increase rapidly above a fluence of 1021 n/cm² as shown in Fig.
a) O₂SCC: In order to grasp the corrosion potential of the material at the area of head and thread in the bolt stagnant region, the calculation on diffusion of dissolved hydrogen and oxygen using by FEM diffusion calculation model and on pH value based upon dissociation of H₃BO₃ at high temperature were conducted, and also data for the kinetics of surface reactions such as exchange current density, Tafel's slope, diffusion limited current density were measured.
b) D/W SCC: From our FEM temperature distribution calculation using by Y -ray heating distribution and heat transfer coefficient of BFBs, the temperature at the surface of BFB shank area is slightly over 343 ℃ of boiling temperature, in case of 120% of Y-ray heating for a 3 loop PWR plant, but is under 343 C in case of best fit Y -ray heating for 2, 3 and 4 loop PWR plants. From our SCC tests under condensed alkaline caused by dry and wet phenomenon using by the model BFBs made of unirradiated 3 16 CW stainless steel, loaded actual stress and added with crevice devised under heat transfer condition at 331 to 332 C in simulated PWR primary water, IGSCC was observed at the shank area of the model BFBs, after 2, 000 hours test.
c) O₂ Low pH type SCC: From our analysis by Nernst- Planck's formula and producing of analytical program about the migration of ion under electro-chemical potential gradient by the experiment of pH changing behavior, and the calculation about diffusion coefficient of H₃BO₃ at high temperature from the electric conductivity, based upon the correlation between the diffusion coefficient of ion and electric conductivity, the water chemistry of bolt stagnant region is decreasing of pH revealed due to the difference of oxygen concentration, but the pH in bolt stagnant region is within neutral, even though lowest pH (about 3) of IGSCC for sensitized 304 stainless steel. It is concluded that there is no possibility of low pH SCC due to oxygen concentration cell caused by residual oxygen.
d) PWSCC: From our accelerated SCC tests by slow strain rate tensile (SSRT) tests in simulated PWR primary water using by the irradiated materials, the effect of test temperature on the threshold fluence for IGSCC is remarkable and the threshold neutron fluence has the tendency to become lower with an increase of test temperature. This tendency looks-like the PWSCC susceptibility of Ni based alloys in simulated PWR primary water.

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