Empirical equations of crack growth rates based on data fitting of neutron irradiated stainless steel under high temperature water simulating boiling water reactor core conditions

Kasahara, Shigeki; Chimi, Yasuhiro; Hata, Kuniki; Fukuya, Koji*; Fujii, Katsuhiko*

Proceedings of 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors (Internet), p.1345 - 1355, 2019/08

This paper describes empirical equation development of crack growth rates (CGR) in consideration of IASCC of neutron irradiated stainless steel to contribute to structural integrity assessment of BWR reactor internals. Empirical equations of CGR (da/dt) were developed based on a formula of da/dt = MK, assuming that "M" and "n" tend to be saturated with increasing neutron fluence. To obtain the empirical equations for normal water chemistry (NWC) and hydrogen water chemistry (HWC) of BWR, a data fitting with least square method was applied to the datasets consisting of F, K and CGR from post irradiation examinations of neutron irradiated stainless steel under simulated NWC and HWC conditions from open literature. As a result, calculated results by the equation for NWC showed good agreement with measured CGR data, meanwhile those for HWC did not. The above difference was seemed to be attributed that CGR data obtained under HWC conditions were scattered extensively.

Estimation of radiocesium dietary intake from time series data of radiocesium concentrations in sewer sludge

Pratama, M. A.; Takahara, Shogo; Munakata, Masahiro; Yoneda, Minoru*

Environment International, 115, p.196 - 204, 2018/06

https://doi.org/10.1016/j.envint.2018.03.009

Effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels

Aoki, So; Kondo, Keietsu; Kaji, Yoshiyuki; Yamamoto, Masahiro

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.663 - 672, 2018/00

https://doi.org/10.1007/978-3-319-68454-3_51

The objective of this study was to clarify effect of long-term thermal aging on SCC initiation susceptibility in low carbon austenitic stainless steels. Specimens used were Type 304L and 316L steels. Both steels were cold rolled to 20% thickness reduction (CW) and then followed by long-term thermal aging at 288C for 14,000 h (LTA). Crevice Bent Beam (CBB) test was carried out to estimate the SCC initiation susceptibility under BWR simulated water condition at high temperature. The present results of the CBB tests showed that 304L CW + LTA exhibited no SCC susceptibility. In contrast, the SCC initiation susceptibility of 316L increased by the combination of cold rolling and long-term thermal aging. To understand these results, evaluation on the changes in microchemistry, microstructure and mechanical properties induced by the CW and LTA treatment has been developed, and their correlation with the SCC initiation susceptibility is discussed.

Evaluation of crack growth rates and microstructures near the crack tip of neutron-irradiated austenitic stainless steels in simulated BWR environment

Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00

https://doi.org/10.1007/978-3-319-68454-3_76

In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at 288C on neutron-irradiated 316L stainless steels (SSs) at 12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at 2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.

Analysis of intergranular cracking in an alloy steel by hydrogen-enhanced decohesion

Yamaguchi, Masatake; Ebihara, Kenichi; Itakura, Mitsuhiro

Proceedings of 2016 International Hydrogen Conference (IHC 2016); Materials Performance in Hydrogen Environments, p.563 - 571, 2017/00

http://doi.org/10.1115/1.861387_ch64

no abstracts in English

Effects of gamma-ray irradiation on spontaneous potential of stainless steel in zeolite-containing diluted artificial seawater

Kato, Chiaki; Sato, Tomonori; Ueno, Fumiyoshi; Yamagishi, Isao

Proceedings of 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1357 - 1374, 2016/05

With respect to the long-term storage of the zeolite-containing spent Cs adsorption vessels used at the Fukushima Daiichi Nuclear Power Station, the corrosion of the vessel material is one of the most important issues. In this study, we performed electrochemical tests on stainless steel specimens in zeolite-containing artificial seawater under gamma-ray irradiation. The spontaneous potential ESP and critical pitting potential VC of the type 316L steel in systems in contact with various zeolites were measured in order to evaluate the corrosion resistance of the steel. In addition, the water sample was analyzed after being irradiated, in order to determine the concentrations of various dissolved oxidants such as oxygen and hydrogen peroxide, which can accelerate the corrosion process. The steady-state rest potential increased with an increase in the dose rate; however, the increase was suppressed in contact with the zeolites. The VC value of the steel when in contact with the zeolites was slightly smaller than the VC value in bulk water; however, the choice of the zeolite used as herschelite, IE96 and IE911 hardly affect the VC value. The concentration of HO in the bulk water under irradiation also increased with the increase in the dose rate. This increase was suppressed in the systems in contact with the zeolites, owing to the decomposition of the HO by the zeolites. A clear relationship was observed between ESP and the HO concentration. As contact with the zeolites caused the increase in ESP under irradiation to be suppressed, it can be concluded that the presence of zeolites in the spent Cs adsorption vessels can reduce the probability of the localized corrosion of the stainless steel in the vessels.

Washout of clay-rich gouge in a pregrouted fault zone and increase of groundwater inflow during tunnel excavation in Neogene siliceous mudstone (Horonobe, Japan)

Ishii, Eiichi; Hashimoto, Yuta; Inagaki, Daisuke*

Proceedings of 10th Asian Regional Conference of International Association for Engineering Geology and the Environment (IAEG ARS 2015) (USB Flash Drive), 4 Pages, 2015/09

This manuscript describes unpredicted inflow which occurred during tunnel excavation for a permeable fault zone with clay-rich gouge although the fault zone was pregrouted. Some observational evidence indicates that the gouge was increasingly washed out into the tunnel through the rockbolts penetrating the gouge and the boundary between the shotcrete and the gouge on the excavation side wall during excavation. The resultant piping/erosion of the gouge probably accelerated groundwater flow from the outer aquifer of the pregrouted zone to the tunnel. After an excavation outcrop exposing the fault zone was temporarily shotcreted, major inflow occurred from the nucleated flow channel in the fault zone, breaking the shotcrete. When a fault zone including such clay-rich gouge is pregrouted, washout of the gouge during tunnel excavation should be fully cared because the gouge itself is not cemented by pre-grouting due to its low-porosity.

Remediation strategy, capping construction and ongoing monitoring for the mill tailings pond, Ningyo-toge Uranium Mine, Japan

Saito, Hiroshi; Taki, Tomihiro

Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 4 Pages, 2013/09

Ningyo-toge Environmental Engineering Center has been managing the Ningyo-toge Uranium Mine. The Mine is subject to environmental remediation to ensure the radiation protection from the exposure pathways to humans in future, and to prevent occurrence of mining pollution. The Yotsugi Mill Tailings Pond is considered as the highest prioritized facility. So far, capping is put on the stable upstream part of the Pond to reduce dose rate, radon exhalation and rainwater precipitation. Prior to its construction, it was decided that the capping are to be multi-layered and natural materials are to be used, and the capping specifications were decided. Also, drainage channels and gabion work are constructed around the capping and at the foot of capping slope, respectively. Monitoring, which will continue for years, has started to confirm the effectiveness of the capping and the data will be utilized for the future remediation of the downstream of the Pond.

Demonstration test results of organic materials' volumetric reduction using bio-ethanol, thermal decomposition and burning

Tagawa, Akihiro; Watanabe, Masahisa

Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 6 Pages, 2013/09

Behaviour of IPG waste forms bearing BaSO as the dominant sludge constituent generated from the treatment of water used for cooling the stricken power reactors

Amamoto, Ippei; Kobayashi, Hidekazu; Yokozawa, Takuma; Yamashita, Teruo; Nagai, Takayuki; Kitamura, Naoto*; Takebe, Hiromichi*; Mitamura, Naoki*; Tsuzuki, Tatsuya*

Proceedings of 15th International Conference on Environmental Remediation and Radioactive Waste Management (ICEM 2013) (CD-ROM), 8 Pages, 2013/09

The great amount of water used for cooling the stricken power reactors at Fukushima Dai-ichi has resulted in accumulation of "remaining water". As the remaining water is subsequently contaminated by FPs, etc., it is necessary to decontaminate it in order to reduce the volume of liquid radioactive waste and to reuse it again for cooling the reactors. Various techniques are being applied to remove FP, etc. and to make stable waste forms. One of the methods using the iron phosphate glass as a medium is being developed to stabilize the strontium-bearing sludge whose main component is BaSO. From the results hitherto, the iron phosphate glass is regarded as a potential medium for the target sludge.