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Takahashi, Rieko*; Taniguchi, Naoki
Zairyo To Kankyo, 73(6), p.153 - 163, 2024/06
Carbon steel is one of the candidate materials for overpacks in geological disposal of high-level radioactive waste, and is known to susceptible to stress corrosion cracking(SCC) depending on the condition in carbonate environment. In order to understand the influence of temperature on the SCC susceptibility of carbon steel, slow strain rate test (SSRT) of rolled steel were performed in NaHCO
aqueous solution with varying temperature in the range of 303-393K for conditions of 0.1-0.5 mol/dm
, which is assumed to be the upper limit of carbonate concentration in groundwater in a geological disposal environment. As the results, no obvious influence of temperature on mechanical properties such as fracture strain ratio and reduction area ratio were observed, but SCC susceptibility based on SCC fracture ratio increased at relatively low temperatures of 303K and 323K. It was suggested that the reason for the higher SCC sensitivity at lower temperatures was due to slower repassivation at lower temperatures. Regarding the type of SCC, intergranular SCC was dominant at low temperatures and tended to transition to intergranular SCC at higher temperatures. Transgranular SCC tended to be observed at lower potentials than those at which intergranular SCC was observed.
Miura, Yasufumi*; Suzuki, Kenji*; Morooka, Satoshi; Shobu, Takahisa
Quantum Beam Science (Internet), 8(1), p.1_1 - 1_14, 2024/03
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Journal of Nuclear Engineering and Radiation Science, 5(3), p.031505_1 - 031505_8, 2019/07
Probabilistic fracture mechanics (PFM) analysis is expected as a rational method for the structural integrity assessment because it can consider the uncertainties of various influence factors and can evaluate the quantitative value such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for the structural integrity assessment of piping welds in nuclear power plants. In the latest few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in the nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus the structural integrity assessment taking account of PWSCC has become important. In this paper, we improved PASCAL-SP for the assessment considering PWSCC by introducing the several analytical functions such as the evaluation models of crack initiation time, crack growth rate and probability of crack detection. By using improved PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were evaluated as numerical examples. We also evaluated the influence of a leak detection and a non-destructive examination on the failure probabilities. On the basis of the numerical results, we concluded that the improved PASCAL-SP is useful for evaluating the failure probability of pipe taking PWSCC into account.
Uchida, Shunsuke; Chimi, Yasuhiro; Kasahara, Shigeki; Hanawa, Satoshi; Okada, Hidetoshi*; Naito, Masanori*; Kojima, Masayoshi*; Kikura, Hiroshige*; Lister, D. H.*
Nuclear Engineering and Design, 341, p.112 - 123, 2019/01
Times Cited Count：7 Percentile：57.46(Nuclear Science & Technology)Improvement of plant reliability based on reliability-centered-maintenance (RCM) is going to be undertaken in NPPs. RCM is supported by risk-based maintenance (RBM). The combination of prediction and inspection is one of the key issues to promote RBM. Early prediction of IGSCC occurrence and its propagation should be confirmed throughout the entire plant systems which should be accomplished by inspections at the target locations followed by timely application of suitable countermeasures. From the inspections, accumulated data will be applied to confirm the accuracy of the code, to tune some uncertainties of the key data for prediction, and then, to increase their accuracy. The synergetic effects of prediction and inspection on application of effective and suitable countermeasures are expected. In the paper, the procedures for the combination of prediction and inspection are introduced.
Yamamoto, Masahiro; Soma, Yasutaka; Igarashi, Takahiro; Ueno, Fumiyoshi
Proceedings of Annual Congress of the European Federation of Corrosion (EUROCORR 2018) (USB Flash Drive), 7 Pages, 2018/09
In order to clarify the SCC behavior of SUS316L under BWR environment, mass transfer inside crevice of SUS316L in high temperature water using various crevice gap samples was investigated. The samples were prepared by put together two SUS316L sheets. Crevice gap differs from 0.005 mm to 0.1 mm. Corrosion tests were conducted in 8 ppm dissolved oxygen (DO) conditions. Surface oxide film was analysed by laser Raman spectroscopy (LRS) after immersion. Numerical simulations were also conducted by using COMSOL Maltiphysics. Diffusion process of DO and the other chemical species were calculated with connected to electrochemical process. Electrical conductivities inside the crevice were 100 times larger than these of outer water. The reason of high conductivity is existence of Fe
ions at the DO depletion crevice.
Chimi, Yasuhiro; Sato, Kenji*; Kasahara, Shigeki; Umehara, Ryuji*; Hanawa, Satoshi
Proceedings of Contribution of Materials Investigations and Operating Experience to Light Water NPPs' Safety, Performance and Reliability (FONTEVRAUD-9) (Internet), 10 Pages, 2018/09
To investigate the influence of Zinc (Zn) injection on primary water stress corrosion cracking (PWSCC) growth behavior, crack growth tests of 10% cold-worked Alloy 600 were performed in simulated primary water environment of pressurized water reactor (PWR) at 320
C with a low-concentration (5-10 ppb) Zn injection under dissolved hydrogen (DH) conditions of 5, 30, and 50 cc/kgH
O. As a result of the crack growth tests, DH-dependence of crack growth rate (CGR) showed a similar tendency to the predicted CGR based on the CGR data without Zn injection, indicating almost no effect of a low-concentration Zn injection on the crack growth behavior. Moreover, the microstructural analyses of oxide films formed inside the crack and on the specimen surface were conducted, and the intake of Zn in the oxides was detected on the specimen surface, but not detected inside the crack. This result was considered to be the cause of no Zn injection effect on the crack growth behavior.
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
Times Cited Count：2 Percentile：55.47(Materials Science, Multidisciplinary)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.
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Recently, cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel based alloy welds in the primary piping of pressurized water reactors. Structural integrity assessments taking PWSCC into account have become important. Probabilistic fracture mechanics (PFM) is expected as one of rational methods for the assessments because it can account for uncertainty of the influencing factors and evaluate the failure probabilities of components. In JAEA, a PFM analysis code PASCAL-SP was developed to evaluate the failure probability of nuclear pipe. This paper details improvement of the PASCAL-SP to evaluate the failure probability taking PWSCC into account. As numerical examples, the failure probabilities for circumferential and axial cracks due to PWSCC are evaluated. Influence of inspection on failure probabilities are evaluated. As the results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability taking PWSCC into account.
Pokor, C.*; Herbelin, A.*; Couvant, T.*; Kaji, Yoshiyuki
NEA/NSC/R(2016)5 (Internet), p.317 - 360, 2017/05
In aged BWR plants, certain locations in the mid-plane of the core shroud experience fluence levels at which the materials become susceptible to irradiation assisted stress corrosion cracking (IASCC). BWRVIP (Boiling Water Reactor Vessel Internals Program) has developed crack growth disposition methodologies for evaluating intergranular stress corrosion cracking (IGSCC) in the internal components of BWRs and the Japan Nuclear Energy Safety organization (JNES) has been conducting a project related to IASCC crack growth rate data as a part of safety research and development study for the aging management and maintenance of the nuclear power plants. Although many investigators proposed prediction models for SCC and IASCC growth rates for austenitic stainless steels and Ni alloys, even more improvements of models are necessary as compared with the detailed experimental results, because these models are still preliminary models.
Hata, Kuniki
Hoshasen Kagaku (Internet), (103), P. 65, 2017/04
no abstracts in English
Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka
Nuclear Engineering and Design, 307, p.411 - 417, 2016/10
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (
), stress intensity factor (
), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.
Chimi, Yasuhiro; Kitsunai, Yuji*; Kasahara, Shigeki; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 475, p.71 - 80, 2016/07
Times Cited Count：11 Percentile：69.28(Materials Science, Multidisciplinary)To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%-2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps.
Kaji, Yoshiyuki; Miwa, Yukio; Tsukada, Takashi
Nihon Kikai Gakkai M&M 2005 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.175 - 176, 2005/12
From results of analysis of plastic deformation behavior at SCC crack tip by EBSP method, the conclusions was obtained as follows: (1)Regarding to basic data for quantitative evaluation of plastic deformation behavior at SCC crack tip, we obtained relationship between plastic deformation and misorientation of grain using SSRT result for SUS316L stainless steel in high temperature water and local plastic strain was able to estimate using this relationship. (2)From experimental results of plastic strain at SCC crack and ductile crack tips using relationship between misorientation and plastic deformation, plastic strain near ductile crack tip was larger than that at SCC crack tip. Plastic deformation region was, however, corresponded to region of about one grain at crack tip for both SCC and ductile cracks.
Kaji, Yoshiyuki; Ugachi, Hirokazu; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
HPR-364, Vol.1 (CD-ROM), 10 Pages, 2005/10
Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this paper, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack propagation and so on, and the present status of in-pile IASCC growth tests using pre-irradiated materials at JMTR.
Ueno, Fumiyoshi; Nagae, Yuji*; Nemoto, Yoshiyuki; Miwa, Yukio; Takaya, Shigeru*; Hoshiya, Taiji*; Tsukada, Takashi; Aoto, Kazumi*; Ishii, Toshimitsu; Omi, Masao; et al.
JAERI-Research 2005-023, 132 Pages, 2005/09
JAERI-Research-2005-023.pdf:33.03MB
JAERI and JNC have started a JAERI-JNC joint research program in fiscal year 2003, which has been aimed for efficient progress and synergistic effect on the research activities in both Institutes. This study has been chosen one of the joint research themes because it has been our common objective in the field of structural materials of FBR and LWR components. The purpose of the study is to clarify damage mechanism of structural materials used under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage. In fiscal year 2004 and 2005, micro-corrosion measurement, electrochemical corrosion test and leakage magnetic flux density measurement apparatuses were developed and equipped in two hot facilities and irradiated and unirradiated crept specimens, irradiated high purity model austenitic stainless alloys were also prepared and applied to this study. These apparatuses and specimens were used for damage evaluation, and these feasibilities for nuclear power plant materials were studied.
Kaji, Yoshiyuki
Proceedings of KNS-AESJ Joint Summer School 2005 for Students and Young Researchers, 2, p.221 - 228, 2005/08
For core internals, the main research items are intergranular stress corrosion cracking (IGSCC) of low carbon stainless steel in core shrouds and primary loop recirculation pipes in boiling water reactor (BWR), and irradiation assisted stress corrosion cracking (IASCC) which is caused by the synergistic effects of neutron and gamma-ray radiation, corrosion by high temperature water, and the residual and/or applied stresses. This paper describes the current status and typical results of fundamental study for mechanistic understanding of IGSCC and IASCC, development of IASCC evaluation technology for BWR plants based on post-irradiation IASCC test data as a part of METI's national project, in-pile IASCC tests.
Department of JMTR
JAERI-Review 2004-029, 100 Pages, 2005/01
JAERI-Review-2004-029.pdf:19.03MB
no abstracts in English
Ugachi, Hirokazu; Kaji, Yoshiyuki; Nakano, Junichi*; Matsui, Yoshinori; Kawamata, Kazuo; Tsukada, Takashi; Nagata, Nobuaki*; Dozaki, Koji*; Takiguchi, Hideki*
Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.319 - 325, 2005/00
Irradiation assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors (LWRs) for a long period. In the research field of IASCC, mainly PIEs for irradiated materials have been carried out, because there are many difficulties on SCC tests under neutron irradiation. Hence as a part of the key techniques for in-pile SCC tests, we have embarked on a development of the test technique to obtain information concerning effects of applied stress level, water chemistry, irradiation conditions, etc. In this conference, we describe the developed several techniques, especially control of loading on specimens, monitoring technique of crack initiation, propagation and water chemistry, and the current status of in-pile SCC tests using thermally sensitized materials at JMTR.
Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro
Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors (CD-ROM), p.311 - 318, 2005/00
Plastic deformation behavior to influence the stress corrosion cracking was studied for the thermally-sensitized and the irradiated type 316LN stainless steel. SSRT was conducted at 573 K in oxygenated water (DO=10ppm) for specimens. Each of the specimens was thermally sensitized at 1033 K for 100 h or irradiated at 473 K to 1 dpa. Between these specimens, the plastic deformation behavior and the IGSCC were compared. For the irradiated specimens, plastic deformation behavior such as the work hardening capability and the maximum stress where IASCC initiated was similar to that of thermally-sensitized specimens in true stress-true strain relation. Moreover, the effect of strain rate on %IGSCC was the same each other. It was suggested from these results that for specimens irradiated around 1 dpa, the initiation mechanism of IASCC was similar to that of IGSCC for thermally-sensitized specimens.
Hoshiya, Taiji*; Ueno, Fumiyoshi; Takaya, Shigeru*; Nagae, Yuji*; Nemoto, Yoshiyuki; Miwa, Yukio; Aoto, Kazumi*; Tsukada, Takashi; Abe, Yasuhiro*; Nakamura, Yasuo*; et al.
JAERI-Research 2004-016, 53 Pages, 2004/10
JAERI-Research-2004-016.pdf:22.07MB
Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Energy Research Institute (JAERI) have started a JNC-JAERI united research program cooperatively in 2003, which has been aimed for efficient progress and synergistic effect on the research activities of both Institutes in order to lead the facing task of unification between JNC and JAERI. This study has been chosen one of the united research themes, and the purpose of it is to clarify damage mechanism of structural materials under irradiation, and then to develop the methods for damage evaluation and detection in earlier stage of progressing process of damage. In fiscal year 2003, magnetic flux density distribution (JNC) and micro-corrosion (JAERI) measurement apparatus were newly developed and equipped in Hot Facilities in two Institutes, respectively. These apparatus were designed and produced in consideration of radiation resistance and remote-controlled operation to equip in hot cells. We will start the study on neutron irradiation damage by employing the two apparatus as the next step.
