Collaborative Laboratories for Advanced Decommissioning Science; National Institute of Maritime, Port and Aviation Technology*
JAEA-Review 2021-049, 67 Pages, 2022/01
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2020. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2020, this report summarizes the research results of the "Research and development of the sample-return technique for fuel debris using the unmanned underwater vehicle" conducted in FY2020. The present study aims to develop a fuel debris sampling device that comprises a neutron detector with radiation resistance and enhanced neutron detection efficiency, an end-effector with powerful cutting and collection capabilities, and a manipulator under the Japan-UK joint research team. We will also develop a fuel debris sampling system that can be mounted on an unmanned vehicle. In addition, we will develop a positioning system to identify the system position, and a technique to project the counting information of optical cameras, sonar, and neutron detectors to be developed in this
Narukawa, Takafumi; Udagawa, Yutaka
Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10
Soler, J. M.*; Meng, S.*; Moreno, L.*; Neretnieks, I.*; Liu, L.*; Keklinen, P.*; Hokr, M.*; ha, J.*; Vetenk, A.*; Reimitz, D.*; et al.
SKB TR-20-17, 71 Pages, 2021/07
Task 9B of the SKB Task Force on Modelling of Groundwater Flow and Transport of Solutes in fractured rock focused on the modelling of experimental results from the LTDE-SD in situ tracer test performed at the sp Hard Rock Laboratory in Sweden. Ten different modelling teams provided results for this exercise, using different concepts and codes. Three main types of modelling approaches were used: (1) analytical solutions to the transport-retention equations, (2) continuum-porous-medium numerical models, and (3) microstructure-based models accounting for small-scale heterogeneity (i.e. mineral grains and microfracture distributions). The modelling by the different teams allowed the comparison of many different model concepts, especially in terms of potential zonations of rock properties (porosity, diffusion, sorption), such as the presence of a disturbed zone at the rock and fracture surface, the potential effects of micro- and cm-scale fractures.
Endo, Takashi*; Tachi, Yukio; Ishidera, Takamitsu; Terashima, Motoki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 20(1), p.9 - 22, 2021/03
Evaluation method of colloid diffusion and filtration in compacted bentonites was developed using dendrimers. Diffusion and filtration behavior of PAMAM dendrimers with the size of 5.77.2nm was investigated by the through-diffusion experiment in bentonite compacted to 0.8 Mg/m and saturated with 0.0050.5mol/L NaCl. Effective diffusivities (De) and filtration ratios (Rf) of dendrimers were determined from the breakthrough curves and the depth profiles in compacted bentonite, respectively. The De values of negatively charged dendrimer increased when porewater salinity increased and dendrimer size decreased as influenced by anion exclusion effect in negatively charged clay surfaces. The Rf values increased when porewater salinity decreased and dendrimer size increased, demonstrating significant fractions of dendrimer were filtered by narrow pores in complex pore networks. These trends consistent with the previous studies emphasize the validity of the evaluation method using dendrimer.
Okubo, Takahiro*; Yamazaki, Akio*; Fukatsu, Yuta; Tachi, Yukio
Microporous and Mesoporous Materials, 313, p.110841_1 - 110841_11, 2021/01
Pore distributions in water-saturated Ca-montmorillonite were investigated using H NMR measurements under various dry densities (0.8 - 1.6 g/cm) and porewater salinity conditions (deionized water, 0.1 and 1 M CaCl), at the temperature range of 233 - 303 K. The volume fractions of the interlayer pore including two and three hydrated layers and the non-interlayer pore in compacted Ca-montmorillonite were quantified by NMR relaxometry including and distribution analysis, and were compared with NMR cryoporometry and X-ray diffractometry. These analysis provided consistent pictures on the pore distributions in compacted Ca-montmorillonite, in contrast to Na-montmorillonite. The main factor affecting the pore distribution in compacted Ca- and Na-montmorillonite is the density, whereas the effect of porewater salinity is relatively smaller. The effect of interlayer cations is also relatively smaller at higher density, although the differences in the pore structures are significant at low density.
Harjo, S.; Kawasaki, Takuro; Grazzi, F.*; Shinohara, Takenao; Tanaka, Manako*
Materialia, 7, p.100377_1 - 100377_9, 2019/09
Takahashi, Hiroaki*; Tachi, Yukio
Applied Clay Science, 168, p.211 - 222, 2019/02
Microstructural and mass transport properties of compacted Na- and Cs-montmorillonites with different swelling properties were investigated by combining 3D microstructure analysis using nanofocus X-ray CT and diffusion measurement of HDO. The X-ray CT observations indicated that macropores in the dry state of compacted Na-montmorillonite are filled with gel phases, and the grain sizes of clay particles shifted toward smaller values through the saturation and swelling processes. By contrast, no gel phase and no decrease in the grain and pore volumes were observed for saturated Cs-montmorillonite. The geometrical factors of the macropores including tortuosity and geometric constrictivity of saturated Cs-montmorillonite determined by the X-ray CT was consistent with the corresponding values derived in the HDO diffusion test. In the case of Na-montmorillonite, the larger differences between the geometric factors evaluated by the X-ray CT and the diffusion tests can be explained by the electrostatic constrictivity factor and the additional geometrical factors in gel phase and interlayer that are smaller than the detection limit of the X-ray CT.
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.
Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 480, p.386 - 392, 2016/11
This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.
Chimi, Yasuhiro; Kitsunai, Yuji*; Kasahara, Shigeki; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 475, p.71 - 80, 2016/07
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.
Taguchi, Tomitsugu; Igawa, Naoki; Yamamoto, Hiroyuki; Shamoto, Shinichi; Jitsukawa, Shiro
Physica E, 28(4), p.431 - 438, 2005/09
Preparation conditions of single-phase SiC nanotubes and C-SiC coaxial nanotubes were investigated. The characterization of single-phase SiC nanotubes and C-SiC coaxial nanotubes were carried out. The SiC nanowires, which were made of the catenated SiC grains of 50-200 nm in diameter, were obtained in carbon nanotubes reacted at 1450 C. The only C-SiC coaxial nanotubes were formed at 1300 C. A few single-phase SiC nantoubes were synthesized at 1200 C for 100 h. More than half number of nanotubes reacted at 1200 C for 100 h were altered to single-phase SiC nantoubes by heat treatment of 600 C for 1h in air since the remained carbon was removed. The energy dispersive X-ray spectroscopy analysis revealed that the atomic ratio of Si to C in single-phase SiC nanotubes was almost 1; these single-phase SiC nanotubes consisted of near-stoichiometric SiC grains.
Taguchi, Tomitsugu; Igawa, Naoki; Miwa, Shuhei*; Wakai, Eiichi; Jitsukawa, Shiro; Snead, L. L.*; Hasegawa, Akira*
Journal of Nuclear Materials, 335(3), p.508 - 514, 2004/12
The microstructure of near-stoichiometric fiber SiC/SiC composites implanted with He and H ions was studied at implantation temperatures of 1000 and 1300 C. The average size of He bubbles in the CVI SiC matrix decreases with increasing concentration of implanted H ions. Moreover, the number density of He bubbles increases with increasing irradiation temperature and amount of implanted H. At the irradiation temperature of 1000 C, He bubbles were mainly formed at grain boundary within the matrix. On the other hand, He bubbles were formed both at grain boundaries and within grains at the irradiation temperature of 1300 C. The average size of He bubbles at grain boundaries was much larger than within the grain. The average size of He bubbles in the fiber was smaller than that in the matrix in all cases.
Shibata, Taiju; Motohashi, Yoshinobu*; Ishihara, Masahiro; Baba, Shinichi; Ito, Tsutomu*; Sawa, Kazuhiro
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai (2004) Koen Rombunshu (No.040-3), p.63 - 64, 2004/09
no abstracts in English
Taguchi, Tomitsugu; Nozawa, Takashi*; Igawa, Naoki; Kato, Yutai*; Jitsukawa, Shiro; Koyama, Akira*; Hinoki, Tatsuya*; Snead, L. L.*
Journal of Nuclear Materials, 329-333(Part1), p.572 - 576, 2004/08
The SiC/SiC composite with SiC/C multi-layer interphase coated on advanced SiC fibers was fabricated by the forced thermal-gradient chemical vapor infiltration (F-CVI) process for improvement in mechanical properties. The SEM and TEM observation verified that SiC/C multi-layer interphase was formed on SiC fibers. The both flexural and tensile strengths of SiC/SiC composite with SiC/C multi-layer interphase were approximately 10 % higher than that with single carbon interphase. The SEM observation on the fracture surface of the composite with SiC/C multi-layer reveals that cylindrical steps around the fiber were formed. The several crack deflections occurred within SiC/C multi-layer interphase. The SiC/C multi-layer applied in this study operated efficiently to improve the mechanical properties.
Nemoto, Yoshiyuki; Hasegawa, Akira*; Sato, Manabu*; Abe, Katsunori*; Hiraoka, Yutaka*
Journal of Nuclear Materials, 324(1), p.62 - 70, 2004/01
In this study, stress-relieved specimens and recrystallized specimens of pure Mo and Mo-Re alloys (Re content=2,4,5,10,13 and 41wt%) were neutron irradiated up to 20dpa at various temperatures (681-1072K). On microstructure observation, sigma phase and chi phase precipitates were observed in all irradiated Mo-Re alloys. Voids were observed in all irradiated specimen, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo-41Re were drastically embrittled after irradiation at 874K or less. From these results, authors discuss about relation between microstructure development and radiation hardening, embrittlement, and propose the most efficient Re content and thermal treatment for Mo-Re alloys to be used under irradiation condition.
Shibata, Taiju; Baba, Shinichi; Yamaji, Masatoshi*; Sumita, Junya; Ishihara, Masahiro
Nihon Kikai Gakkai M&M 2004 Zairyo Rikigaku Kanfarensu Koen Rombunshu, p.407 - 408, 2004/00
no abstracts in English
Hanawa, Satoshi; Ishihara, Masahiro; Shibata, Taiju
Transactions of 17th International Conference on Structural Mechanics in Reactor Technology (SMiRT-17) (CD-ROM), 6 Pages, 2003/08
From a viewpoint of advanced design method of graphite components, it is important to apply the realistic fracture model in the design method. The applicability of the microstructure based brittle fracture model under multiaxial stress condition was, therefore, investigated. The fracture model is possible to treat grain size as well as pore size with fracture mechanics approach taking account of the crystal structure of the graphite. The model was applied to the biaxial strength prediction of near isotropic nuclear graphite using grain/pore related microstructural parameters. Prediction results were compared with biaxial strength data obtained by simultaneous loadings of inner pressure and longitudinal load with thin-walled cylindrical specimen. From this study, it was found that the fracture model predicted fairly good not only mean strength but also strength distribution under biaxial stress condition, and it was concluded that the microstructure based brittle fracture model would be applicable as the advanced design method.
Yamashita, Toshiyuki; Kuramoto, Kenichi; Shirasu, Noriko; Nakano, Yoshihiro; Akie, Hiroshi; Nagashima, Hisao; Kimura, Yasuhiko; Omichi, Toshihiko*
Journal of Nuclear Materials, 320(1-2), p.126 - 132, 2003/07
Two irradiation tests on the rock-like oxide (ROX) fuels, small disk-shape fuel targets and pellet-type fuels, were performed in order to clarify in-pile irradiation stabilities. Swelling, fractional fission gas release (FGR) and phase change were examined by puncture test, profilometry and ceramography. YSZ single-phase fuel showed an excellent irradiation behavior, ie. low fission gas release (less than 3%), negligible swelling and no appreciable restructuring. The particle dispersed fuels showed lower swelling and higher fission gas release than those of mechanically blended fuels. Spinel decomposition and subsequence restructuring in the spinel matrix fuels was observed for the first time in the present investigation. It would be possible to reduce the FGR of the spinel matrix fuels, if the maximum fuels temperature is limited below 1700 K where neither spinel decomposition nor restructuring was observed. Damaged area of spinel matrix due to fission fragment irradiation seemed to be confined to thin layers around the surface of YSZ particles.
Nakamura, Takehiko; Sasajima, Hideo; Yamashita, Toshiyuki; Uetsuka, Hiroshi
Journal of Nuclear Materials, 319, p.95 - 101, 2003/06
Pulse irradiation tests under simulated RIA conditions were performed with three types of ROX fuels. Single phase YSZ, homogeneous mixture of YSZ/spinel and YSZ particle dispersed in spinel type ROX fuels were pulse irradiated in the Nuclear Safety Research Reactor (NSRR). Mode and threshold of the fuel failure including its consequences were investigated under the RIA conditions. The fuel failure occurred in a burst type mode in all the three types of ROX fuel tests with considerable fuel melting. Even though the mode was quite different from those of UO fuel, failure threshold enthalpies of the ROX fuels were close to that of UO fuel at about 10 GJ m. The consequence of the failure of the ROX fuels was different from the one of UO fuel, because molten fuel dispersal occurred at lower enthalpies in the ROX fuel tests. Change of the fuel structure and material interaction in the transient heating conditions were examined through optical and secondary electron microscopy, and electron probe micro analysis.
Jitsukawa, Shiro; Tamura, Manabu*; Van der Schaaf, B.*; Klueh, R. L.*; Alamo, A.*; Petersen, C.*; Schirra, M.*; Spaetig, P.*; Odette, G. R.*; Tavassoli, A. A.*; et al.
Journal of Nuclear Materials, 307-311(Part1), p.179 - 186, 2002/12
Reduced activation ferritic/martensitic steel is the primary candidate structural material for ITER Test Blanket Modules and DEMOnstration fusion reactor because of its excellent dimensional stability under irradiation and lower residual activity as compared with the Ni bearing steels such as the austenitic stainless steels. In this paper, microstructural features, tensile, fracture toughness, creep and fatigue properties of a reduced activation martensitic steel F82H (8Cr-2W-0.04Ta-0.1C) are reported before and after irradiation, in addition to the design concept used for development of this alloy. A large number of collaborative test results including those generated under the IEA working group implementing agreements are collected and are used to evaluate the feasibility of use of F82H steel as one of the reference alloys. The effect of metallurgical variables on the irradiation hardening is reviewed and compared with the results obtained from irradiation experiments.