Effect of dissolved oxygen concentration on dynamic strain aging and stress corrosion cracking of SUS304 stainless steel under high temperature pressurized water

Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*

The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01

https://doi.org/10.1063/5.0192888

Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.

Numerical simulation of oxygen infusion into desaturation resulting from artificial openings in sedimentary formations

Miyakawa, Kazuya; Aoyagi, Kazuhei; Akaki, Toshifumi*; Yamamoto, Hajime*

Dai-15-Kai Iwa No Rikigaku Kokunai Shimpojiumu Koen Rombunshu (Internet), p.609 - 614, 2021/01

Desaturation is expected due to excavation of an underground repository, especially in the newly created fractures zone (EDZ). During the construction and operation of facilities, the air in the gallery infuses into the rock around the gallery though the excavation affected area and causes oxidation of host rock and groundwater, which increase nuclide mobilities. In the Horonobe underground research laboratory (HURL), which is excavated in the Neogene sedimentary formations, no pyrite dissolution or precipitation of calcium sulfates was found from the cores drilled in the rock around the gallery. The reason for no oxidation is estimated that the release of dissolved gases from groundwater due to pressure decrease flows against the air infusion. In this research, the mechanism of O intrusion into the rock was investigated by numerical multiphase flow simulation considering advection and diffusion of groundwater and gases. In the simulation, only Darcy's and Henry's laws were considered, that is, chemical reaction related to oxidation was not handled. The effects of dissolved gas and rock permeability on O infusion into the rock were almost identical. Decreasing humidity with relatively low permeability leads to extensive accumulation of O into the EDZ even though with a relatively large amount of dissolved gas. In the HURL, the shotcrete attenuates O concentration and keeps 100% humidity at the boundary of the gallery wall, which inhibits O infusion. Without the shotcrete, humidity at the gallery wall decreases according to seasonal changes and ventilation, which promotes O intrusion into the EDZ but the chemical reaction related to O buffering such as pyrite oxidation consumes O.

Report of Examination of the Samples from Core Shroud (2F3-H6a) at Fukushima Dai-ni Nuclear Power Station Unit-3 (Contract Research)

The Working Team for Examination Operation of Samples From Core Shroud at Fukushima Dai-ni Unit-3

JAERI-Tech 2004-044, 92 Pages, 2004/05

JAERI-Tech-2004-044.pdf:15.18MB

The present examination has been performed with the objective to ensure the transparency of the examination as the third-party organization by providing technical basis for identifying the causes of cracking through examination of the sample taken from the cracked region of outer H6a welding portion of the core shroud at Fukushima Dai-ni Nuclear Power Station Unit-3, which was a part of sample stored in the Nippon Nuclear Fuel Development Co., Ltd. in the examination of Tokyo Electric Power Company in 2001. The present examination of the sample was conducted at the post irradiation examination facilities of JAERI. The following findings were obtained from the result of the present examination. (1)Three cracks were observed at the portion 3 to 9mm apart from the weld metal and the maximum depth was about 8mm. (2)Intergranular cracking was observed in almost whole fracture surface. The transgranular cracking was partially observed within the depth of about 300m from the surface. (3)Hardening layer over Hv400 at its maximum was found from the surface to the depth of about 500m. Based on the examination results concerning presence of tensile residual stress by welding and relatively high dissolved oxygen contents in core coolant, it is concluded that the cracks were mainly initiated in the hardening layer by transgranular stress corrosion cracking and propagated along the grain boundaries.

Report of Examination of the Sample from Core Shroud (K1-H4) at the Kashiwazaki-Kariwa Nuclear Power Station Unit-1 (Contract research)

The Working Team for Examination of the Sample from Core Shrouds and Primary Loop Recirculation Pipi

JAERI-Tech 2004-011, 64 Pages, 2004/02

JAERI-Tech-2004-011.pdf:14.65MB

At the Kashiwazaki-Kariwa Nuclear Power Station Unit-1 of the TEPCO, cracks were confirmed at the weld joint (H4) in the middle of core shroud, by the visual inspection test for the weld joint of core shroud during the 13th periodic examination by a direction of the Nuclear and Industrial Agency. TEPCO has conducted a material examination with NFD on the specimen including cracks sampled from the core shroud. The present research has been performed with the objective to independently investigate and evaluate the materials by jointly attending the examination with NFD from the planning stage, receiving the final data given by the examination and providing JAERI's own evaluation report as a third-party organization for assuring the transparency. As a result, the consideration of residual stress induced with welding process and dissolved oxygen concentration in core cooling water, it was concluded that the cracks were initiated by SCC and propagated three-dimensionally through grains, and some cracks reached weld metal.

Development of analytical method and study about microstructure of oxide films on stainless steel

Nemoto, Yoshiyuki; Miwa, Yukio; Kikuchi, Masahiko; Kaji, Yoshiyuki; Tsukada, Takashi; Tsuji, Hirokazu

Journal of Nuclear Science and Technology, 39(9), p.996 - 1001, 2002/09

https://doi.org/10.3327/jnst.39.996

Surface morphology of oxidized stainless steel was evaluated using atomic force microscope (AFM) and scanning electron microscope (FE-SEM). Cross-sectional morphology of oxide layer on the specimens was evaluated using FE-SEM after fabrication. Focused ion beam (FIB) technique was applied to fabricate thin film samples of oxide films, which were used for microstructure observation by transmission electron microscope (FE-TEM), and microscopic chemical analysis by energy dispersed X-ray spectrometer (EDS). These preparations and observations were successful, and microstructure and chemical composition of oxide films were evaluated on nanometer scale. Effects of silicon (Si) doping and dissolved oxygen (DO) content in water for oxide layer formation are discussed.

Neptunium migration involving oxidation-reduction reactions in engineered barriers

J.Ahn*;

Proc. of the 3rd Int. Symp. on Advanced Nuclear Energy Research; Global Environment and Nuclear Energy, 10 Pages, 1991/00

no abstracts in English

Effect of grain refinement on dynamic strain aging in SUS304L stainless steel under high temperature pressurized water

Hirota, Noriaki; Kondo, Keietsu; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Ide, Hiroshi; Tsuchiya, Kunihiko; Kobayashi, Yoshinao*

no journal, , 

Dynamic strain aging (DSA) has been identified in shrouds of boiling water reactors and recirculation system piping of pressurized water reactors in the nuclear field. This phenomenon increases the work hardening rate of the material and causes a reduction in ductility. Rodriguez reported that using stainless steel, this work hardening increases with grain refinement, making DSA more likely to occur. The objective of this study is to evaluate the effect of grain refinement on DSA in a high temperature pressurized water (HTPW) simulating nuclear reactor environment utilizing ultrafine grained SUS304L (UFGS). UFGS was heat treated to adjust the grain size from 0.59 m to 68.6 m, and Hall-Petch relationship for 0.2 % yield stress was arranged. The k values obtained in this study were almost the same as the reference values previously obtained for SUS304L. Regarding the effect of grain size on fracture strain, a comparison of fracture strain between tensile test under air and slow strain rate test (SSRT) under 598 K / 15 MPa at dissolved oxygen 1 ppb showed that the fracture strain was lower than that under air as the grain size became coarser. The micrograph after fracture in a HTPW showed that ductile fracture surfaces were observed for materials with grain sizes less than 28.4 m. However, when the grain size coarsened to 68.6 m, more than half of all fracture surfaces were brittle fractured. For the material with a grain size of 0.59 m under HTPW, many correspondence grain boundaries of {111}/3 boundaries were observed in the fracture cross-section of the sample. But these distributions were rarely observed when the grain size was coarsened to 68.6 m. Therefore, the suppression of crack propagation by DSA to the fine grains in a HTPW can be attributed to the relaxation of dislocation accumulation by the {111}/3 boundaries.