Oda, Chie; Kawama, Daisuke*; Shimizu, Hiroyuki*; Benbow, S. J.*; Hirano, Fumio; Takayama, Yusuke; Takase, Hiroyasu*; Mihara, Morihiro; Honda, Akira
Journal of Advanced Concrete Technology, 19(10), p.1075 - 1087, 2021/10
Concrete in a transuranic (TRU) waste repository is considered a suitable material to ensure safety, provide structural integrity and retard radionuclide migration after the waste containers fail. In the current study, coupling between chemical, mass-transport and mechanical, so-called non-linear processes that control concrete degradation and crack development were investigated by coupled numerical models. Application of such coupled numerical models allows identification of the dominant non-linear processes that will control long-term concrete degradation and crack development in a TRU waste repository.
Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
Mishima, Seiki*; Ogata, Sho*; Inui, Toru*; Yasuhara, Hideaki*; Kishida, Kiyoshi*; Aoyagi, Kazuhei
Dai-15-Kai Iwa No Rikigaku Kokunai Shimpojiumu Koen Rombunshu (Internet), p.215 - 220, 2021/01
When the durability of the geological repository of high-level radioactive waste is evaluated, understanding the cracking behavior within the crystalline/sedimentary rocks during excavation of waste disposal cavities is important. In this study, we performed a numerical analysis that expressed the tunnel excavation carried out 350 m underground at the Horonobe Underground Research Center of the Japan Atomic Energy Agency. Simulated results are agreement with actual trends of fracture propagation, and the measured horizontal convergence of the tunnel was reproduced by the numerical analysis relatively well.
Okano, Aoi*; Kimoto, Kazushi*; Matsui, Hiroya
Dai-15-Kai Iwa No Rikigaku Kokunai Shimpojiumu Koen Rombunshu (Internet), p.633 - 636, 2021/01
This study evaluates the acoustic anisotropy of granite using surface waves. It is well-known that granite shows acoustic anisotropy due to preferentially oriented microcracks. Therefore it may be possible to gain information on the microcracks from the measurement of the acoustic anisotropy. In the conventional rock core elastic wave test, acoustic anisotropy has been evaluated by the ultrasonic transmission test. However, it is difficult to apply this method to field measurement and irregularly-shaped specimens. Therefore, in this study, we attempted to evaluate the acoustic anisotropy of granite using surface waves. By this method, the acoustic anisotropy was evaluated based on the changes in the surface wave amplitude, velocity, and frequency when the transmission direction was varied stepwise at a constant angle. As a result, the proposed surface wave technique evaluated acoustic anisotropy successfully. Furthermore, it was found that the acoustic anisotropy emerges because the microcracks change the apparent rigidity of the granite specimen.
Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(5), p.051501_1 - 051501_10, 2020/10
JAEA-Review 2020-006, 261 Pages, 2020/09
A literature review was conducted on the increase in surface area of vitrified products of HLW due to the fracturing caused by cooling during glass pouring process and by mechanical impact, from the perspective of a parameter of the radionuclide release model in the performance assessment of geological disposal system studied overseas. The review was focused on the value of surface area increase factor set as a parameter in the model, the experimental work to evaluate an increase in surface area, and how the parameters on surface area were determined based on the experimental results. The surface area obtained from the experiments executed in Japan was also discussed in comparison with the overseas studies. On the basis of the investigation, the effects of various conditions on the surface area were studied, such as a diameter of vitrified product, cooling condition during and after the glass pouring, impact on vitrified products during their handling, environment after the closure of disposal facility, and others. The causes of fracturing are associated with the phenomena or events in the waste management process such as production, transport, storage, and disposal. The surface area increase factors set in the nuclide release model of the glass and their bases were reviewed. In addition, the measured values and the experimental methods for surface increase factors published so far were compared. Accordingly, the methods for measuring surface area as the bases were identified for these factors set in the models. The causes of fracturing and features of these factors were studied with respect to the relation with the waste management process. The results from the review and assessment can contribute to the expanding the knowledge for the conservative and realistic application of these factors to performance assessment, and to the developing and upgrading of safety case as a consequence.
Shimodaira, Masaki; Tobita, Toru; Takamizawa, Hisashi; Katsuyama, Jinya; Hanawa, Satoshi
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 7 Pages, 2020/08
In JEAC 4206 which prescribes the methodology for assessing the structural integrity of reactor pressure vessels (RPVs), an under-clad crack (UCC) at the inner surface of RPV is postulated, and it is required that the fracture toughness of RPV steels is higher than stress intensity factor for at the crack tip during the pressurized thermal shock event. In the present study, to investigate the effect of cladding on the fracture toughness, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for an RPV steel containing an UCC or a surface crack, and the constraint effect for UCC was also discussed. As the result, we found that the fracture toughness for UCC was considerably higher than that for surface crack. On the other hand, the FEAs showed that the cladding decreased the constraint effect for UCC.
Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06
Ichikawa, Yasuaki*; Kimoto, Kazushi*; Matsui, Hiroya
JAEA-Research 2019-005, 32 Pages, 2019/10
It is important to evaluate the mechanical stability around the geological repository for high-level radioactive waste, during not only the design, construction and operation phases, but also the post-closure period over several millennia. The rock mass around the tunnels could be deformed in response to time dependent behaviors such as creep and stress relaxation. Therefore, this study has started as a joint research with Okayama University from 2016. This report summarize the results of the joint research performed in fiscal year 2017 and 2018. Based on the research results obtained in fiscal year 2016, automatic measurement system was developed, which can collect very large data on surface elastic wave propagation in a short time, also the applicability of various kinds of parameters derived from measured elastic wave data was examined.
Dostl, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; Amaya, Masaki; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*
Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10
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 320C with a low-concentration (5-10 ppb) Zn injection under dissolved hydrogen (DH) conditions of 5, 30, and 50 cc/kgHO. 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
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.
Wakai, Takashi; Kobayashi, Sumio; Kato, Shoichi; Ando, Masanori; Takasho, Hideki*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
This paper describes a thermal fatigue test on a structural model with a dissimilar welded joint. In the present design of JSFR, there may be dissimilar welded joints between ferritic and austenitic steels especially in IHX and SG. Creep-fatigue is one of the most important failure modes in JSFR components. However, the creep-fatigue damage evaluation method has not been established for dissimilar welded joint. To investigate the evaluation method, structural test will be needed for verification. Therefore, a thermal fatigue test on a thick-wall cylinder with a circumferential dissimilar welded joint between Mod.9Cr-1Mo steel and 304SS was performed. Since the coefficients of thermal expansion of these steels were significantly different, buttering layer of Ni base alloy was installed between them. After the completion of the test, deep cracks were observed at the HAZ in 304SS, as well as at HAZ in Mod.9Cr-1Mo steel. There were many tiny surface cracks in BM of 304SS. According to the fatigue damage evaluation based on the finite element analysis results, the largest fatigue damage was calculated at HAZ in 304SS. Large fatigue damage was also estimated at BM of 304SS. Fatigue cracks were observed at HAZ and BM of 304SS in the test, so that analytical results are in a good agreement with the observations. However, though relatively small fatigue damage was estimated at HAZ in Mod.9Cr-1Mo steel, deep fatigue cracks were observed in the test. To identify the cause of such a discrepancy between the test and calculations, we performed a series of finite element analyses. Some metallurgical investigations were also performed.
Lu, K.; Li, Y.
AIMS Materials Science, 4(2), p.439 - 451, 2017/03
Ichikawa, Yasuaki*; Kimoto, Kazushi*; Matsui, Hiroya; Kuwabara, Kazumichi; Ozaki, Yusuke
JAEA-Research 2016-018, 23 Pages, 2016/12
It is important to evaluate the stability of a repository for high-level radioactive waste not only during the design, construction and operation phases, but also during the post-closure period, for time frames likely exceeding several millennia or longer. The rock mass around the tunnels could be deformed through time in response to time dependent behavior. On the other hand, it was revealed that the chemical reaction of groundwater in a rock had an influence on the long-term behavior. An evaluation of the microcracks to have an influence on this mechanical and chemical coupled phenomena should be worked on chiefly. In fiscal year 2015, using a laser Doppler vibrometer that extends a frequency band up to 20 MHz, and measuring the surface wave transmitted through the granite specimens were estimated group velocity. As a result, group velocity until 100 kHz 500 kHz, revealed that tends to decrease while vibrating. The group speed estimate from a group delay was shown to be easier than the estimate by wave number - frequency spectrum. This is because in order to improve reliability, the estimated frequency band is by using a spatially averaged waveform. As a result obtained, in the case of the modeling by the viscoelastic theory of the granite and a microcrack nondestructiveness evaluation, it is thought that it is useful information in the future. In order to use the knowledge of this study, there is a need to clarify the correspondence between the microscopic properties of the medium such as a crack and crystal grain and the change of the group velocity.
Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka
Nuclear Engineering and Design, 307, p.411 - 417, 2016/10
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.
Kato, Chiaki; Ishijima, Yasuhiro; Ueno, Fumiyoshi; Yamamoto, Masahiro
Journal of Nuclear Science and Technology, 53(9), p.1371 - 1379, 2016/09
The effects of crystal textures and the potentials in the anodic oxidation of zirconium in a boiling nitric acid solution were investigated to study the stress corrosion cracking of zirconium in nitric acid solutions. The growth of the zirconium oxide film dramatically changed depending on the applied potential at a closed depassivation potential (1.47 V vs. SSE). At 1.5 V, the zirconium oxide film rapidly grows, and its growth exhibits cyclic oxidation kinetics in accordance with a nearly cubic rate law. The zirconium oxide film grows according to the quantity of electric charge, and the growth rate does not depend on the crystal texture in the pretransition region before the cyclic oxidation kinetics. However, the growth and cracking under the thick oxide film depend on the crystal texture in the transition region. On the normal direction side, the oxide film thickness decreases on average since some areas of the thick oxide film are separated from the specimen surface owing to the cracks in the thick oxide. On the rolling direction side, cracks are found under the thick oxide film, which deeply propagate along the RD without an external stress. The cracks under the thick oxide film propagate to the center of the oxide layer. The cracks in the oxide layer propagate in the (0002)Zr plane in the zirconium matrix. The oxide layer consists of string-like zirconium oxide and zirconium hydride. The string-like zirconium oxide contains orthorhombic ZrO in addition to monoclinic ZrO. As one assumption for the mechanism of crack initiation and propagation without an external stress, it is considered that the oxidizing zirconium hydrides precipitate in the (0002)Zr and then the phase transformation from orthorhombic ZrO to monoclinic ZrO in the oxide layer causes the crack propagation in the (0002) plane.
Lu, K.; Katsuyama, Jinya; Li, Y.
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 8 Pages, 2016/07
Ichikawa, Yasuaki*; Kimoto, Kazushi*; Sato, Toshinori; Kuwabara, Kazumichi; Takayama, Yusuke
JAEA-Research 2015-025, 31 Pages, 2016/03
It is important to evaluate the stability of a repository for high-level radioactive waste not only during the design, construction and operation phases, but also during the post-closure period, for time frames likely exceeding several millennia or longer. The rock mass around the tunnels could be deformed through time in response to time dependent behavior. On the other hand, it was revealed that the chemical reaction of groundwater in a rock had an influence on the long-term behavior. An evaluation of the microcracks to have an influence on these mechanical and chemical coupled phenomena should be worked on chiefly. In fiscal year 2014, this study performed numerical analysis to examine the supersonic scattering attenuation decrement behavior in the crystalline rock and a measurement sequentially last year. The measurement of the head and surface waves were carried out. As a result, group speed was provided. On the other hand, the spread scattering analysis of the elastic wave by the FDTD (Finite Difference Time-Domain) method made a numerical analysis. However, a laboratory finding is different from expectation of the simulation, and crystal anisotropic influence of a microcrack and rock-forming minerals is thought about as a cause of this estrangement. Therefore it was revealed that it was necessary to examine these two points of influence more in future.
Li, Y.; Hasegawa, Kunio; Katsumata, Genshichiro; Osakabe, Kazuya*; Okada, Hiroshi*
Journal of Pressure Vessel Technology, 137(5), p.051207_1 - 051207_8, 2015/10
A number of surface cracks with large aspect ratio have been detected in components of nuclear power plants in recent years. The depths of these cracks are even larger than the half of crack lengths. However, the solutions of the stress intensity factor were not provided for semi-elliptical surface cracks with large aspect ratio in the current fitness-for-service codes. In this study, in order to conduct integrity assessment for cracked components, the solutions of the stress intensity factor were calculated using finite element analysis for semi-elliptical surface cracks with large aspect ratio in plates. Solutions were provided at both the deepest and the surface points of the surface cracks. Some of solutions were compared with the available existing results. As the result, it was concluded that the solutions proposed in this paper are applicable in engineering applications.