Oxidation resistance and stress corrosion cracking susceptibility of 308L and 309L stainless steel cladding layers in simulated PWR primary water

CORROSION ◽  
10.5006/3699 ◽  
2021 ◽  
Author(s):  
Tongming Cui ◽  
Qi Xiong ◽  
Jiarong Ma ◽  
Kun Zhang ◽  
Zhanpeng LU ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Oxidation Resistance ◽  
Layer Structure ◽  
Initiation Site ◽  
Pressurized Water Reactor ◽  
Internal Oxide ◽  
Pressurized Water ◽  
The Matrix ◽  
Primary Water ◽  
Cladding Layers

Exposure and slow strain rate tensile (SSRT) tests were conducted in a simulated pressurized water reactor (PWR) primary water to investigate the oxidation resistance and SCC susceptibility of 308L and 309L stainless steel (SS) cladding layers. A double-layer structure oxide layer grown on 308L SS and 309L SS contained the Cr-enriched nanocrystalline internal layer and the Fe-enriched spinel oxide in the external layer. Ni-enrichment at the matrix/oxide (M/O) boundary was observed. The internal oxide film on 309L SS was thicker and had a lower Cr content than that on 308L SS. Preferential dissolution of inclusions led to pits on 308L SS and 309L SS surfaces during the exposure tests. More inclusions in 309L would decrease its SCC resistance due to the pits can act as the SCC initiation site. 308L SS had a lower susceptibility of SCC than 309L SS in PWR primary water. Lower ferrite content, higher strength/hardness reduced the oxidation and SCC resistance of 309L SS cladding. The effect of ferrite on oxidation and SCC of the SS claddings was discussed.

Effects of Dissolved Hydrogen on Crack Growth Rate of Warm-Rolled 316L Austenitic Stainless Steel in Primary Water Condition

2015 ◽  
Author(s):  
Kyoung Joon Choi ◽  
Seung Chang Yoo ◽  
Taeho Kim ◽  
Seong Sik Hwang ◽  
Min Jae Choi ◽  
...  
Keyword(s):  
Growth Rate ◽  
Stainless Steel ◽  
Crack Growth Rate ◽  
Crack Growth ◽  
Hydrogen Concentration ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Condition ◽  
Primary Water ◽  
Dissolved Hydrogen

With the extension of pressurized water reactor’s design life or continued operation, more careful study on the integrity of the internal structures needs to be pursued. In this study, warm-rolling and heat-treatment were applied to 316L stainless steel, in order to simulate the effect of radiation damage such as hardening and radiation-induced grain boundary segregation. And, the crack growth rate testing under constant load condition was performed in the primary water conditions of a pressurized water reactor. Also, in order to investigate the effect of dissolved hydrogen on the crack growth, the dissolved hydrogen concentration was varied between 30 to 50 cc/kg in simulated primary water condition of a pressurized water reactor. The warm-rolled specimens showed the higher crack growth rate than as-received one. Also, the crack growth rate increased as the dissolved hydrogen concentration increases.

Effectiveness of Stainless Steel Buffer Layer to Address Hot Cracking During Weld Overlay Repair of Dissimilar Metal Alloy 82/182 Welds With Stainless Steel Piping

2008 ◽  
Author(s):  
Nathaniel G. Cofie ◽  
Richard E. Smith ◽  
Richard L. Bax ◽  
Christopher S. Lohse ◽  
Bill Hermanns ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Weld Metal ◽  
Buffer Layer ◽  
Hot Cracking ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Primary Water ◽  
The Impact

Many pressurized water reactor (PWR) plants have used weld overlays to mitigate the pressurizer dissimilar metal welds that are susceptible to primary water stress corrosion cracking (PWSCC). These configurations typically consist of SA-508 Class 2 low alloy steel welded to a stainless steel safe end by Alloy 82/182 weld metal. The overlay weld metal is typically Alloy 52M. In a few cases, solidification cracking (hot cracking) has been observed on the stainless steel portion of the configuration when the first weld overlay layer is deposited. To overcome this problem, a process consisting of deposition of ER308L or ER309L stainless steel buffer layer for the first layer in conjunction with a low Power Ratio welding procedure has been developed and applied successfully. The Alloy 52M weld overlay is then deposited after the buffer layer. This paper discusses the causes of the hot cracking and test programs to develop the parameters for the welding of the buffer layer and subsequent weld overlay layers. In addition the results of analysis performed to determine the impact of the buffer layer on the post weld overlay residual stresses are also discussed.

scholarly journals Grain boundary oxidation of proton-irradiated nuclear grade stainless steel in simulated primary water of pressurized water reactor

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Ping Deng ◽  
Qunjia Peng ◽  
En-Hou Han
Keyword(s):  
Stainless Steel ◽  
Grain Boundary ◽  
Atom Probe ◽  
Nuclear Grade ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oxide Precursor ◽  
Transmission Electron ◽  
Primary Water

AbstractGrain boundary (GB) oxidation of proton-irradiated 304 nuclear grade stainless steel in primary water of pressurized water reactor was investigated. The investigation was conducted by studying microstructure of the oxide and oxide precursor formed at GB on an "atom-by-atom" basis by a combination of atom-probe tomography and transmission electron microscope. The results revealed that increasing irradiation dose promoted the GB oxidation, in correspondence with a different oxide and oxide precursor formed at the GB. Correlation of the oxide and oxide precursor with the GB oxidation behavior has been discussed in detail.

Application of Magnetic Stir Welding to Dissimilar Metal Structural Weld Overlay

2008 ◽  
Author(s):  
Takeshi Yoshida ◽  
Takaaki Matsuoka ◽  
Yuta Uchida ◽  
Takashi Hirano
Keyword(s):  
Stainless Steel ◽  
Molten Pool ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Lack Of Fusion ◽  
Steel Base ◽  
Single Bead ◽  
Primary Water

Alloy 600 and associated welds, Alloy 82/182 of the Pressurized Water Reactor (PWR) plants have been known as being susceptible to the Primary Water Stress Corrosion Cracking (PWSCC). Dissimilar metal (DM) piping butt welds were welded with Alloy 82/182. As one of the mitigation techniques of the PWSCC, Structural Weld Overlay (SWOL) has been applied to the DM welds, but it has tendency to occur weld cracks on the first layer. One of the reasons of the weld cracks is the sulfur which is highly contained in stainless steel base metals, because old stainless steels would contain higher sulfur (e.g. 0.02%) than later ones. In response to this situation, Magnetic Stir Welding (MSW) was proposed to apply for the first layer of SWOL, and tested to evaluate its weldabilities. MSW has been developed for several years, and it is generally known that MSW has characteristics to improve a heat transfer in the molten pool, so that it could reduce a dilution. The purpose of this study is to evaluate weldabilities of MSW for welding Alloy 52 and/or Alloy 52M as filler metal on high sulfur contained stainless steel pipe. Single bead tests and all position welding tests were conducted. As a result of this study, MSW can prevent from occurring weld cracks and lack of fusion due to stirring effects of the molten pool. Therefore, SWOL can be welded without weld cracks on the first layer by applying MSW, even though the stainless steel base metal contains relatively high sulfur. In addition, MSW can weld at high wire supply rate because of prevention of lack of fusion. So it could improve weld efficiency.

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