Characterizing the reactivation mechanisms of coseismic surface ruptures associated with the 2011 Mw 6.7 Fukushima-ken Hamadori earthquake in Japan through borehole hydromechanical testing

Guglielmi, Y.*; Aoki, Kazuhiro; Cook, P.*; Soom, F.*; Cappa, F.*; Tanaka, Yukumo

Tectonophysics, 819, p.229084_1 - 229084_22, 2021/11

https://doi.org/10.1016/j.tecto.2021.229084

The reactivation mechanisms of coseismic surface ruptures associated with the 2011 Mw 6.7 Fukushima-ken Hamadori earthquake in Japan are investigated using in-situ controlled hydraulic injections in subsurface boreholes. Two fault segments were selected for reactivation studies, one across a coseismic rupture, the Shionohira site, and one across a non-coseismically ruptured segment, the Minakami-kita site. A series of water injections in sealed sections of boreholes set across the fault progressively bring the fault to rupture by a step-by-step decrease of the effective normal stress clamping the fault. While the fault is rupturing during these hydraulic stimulations, borehole displacements, fluid pressure and injection flowrate are continuously monitored. Then, the tests were analyzed using fully coupled hydromechanical modeling. The model was calibrated on field data, and a parametric study was conducted to examine the modes of fault reactivation. Coseismic surface rupture of the Shionohira fault showed a pure dilatant slip response to hydraulic tests, while the tectonically un-activated Itozawa fault (South) indicated a complex hybrid response to tests related to both a higher frictional and cohesive strengths of the fault. The analysis of the induced Shionohira slip event showed that it is reasonably modeled as a Coulomb rupture with an eventual dependency of friction on slip velocity, in good accordance with laboratory-derived rate-and-state friction data on the Shinohira gouge samples. In contrast, the Itozawa fault reactivation mechanism appears dominated by tensile failure with limited Coulomb shear failure. Thus, the applied protocol proves to be able to isolate significant differences in fault physical properties and rupture mechanisms between two segments of the same fault system, opening perspectives to better assess near-surface rupture effects, and therefore the safety of structures subject to large earthquakes.

Evaluation of tritium release into primary coolant for research and testing reactors

Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*

Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01

https://doi.org/10.1080/00223131.2020.1791760

The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.

Potential for remote controllable systematization of the method of testing reinforced concrete using guided-wave on rebar

Furusawa, Akinori; Nishimura, Akihiko; Takenaka, Yusuke; Muramatsu, Toshiharu

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

The aim of this work presented here is to demonstrate the potential of our method for remote controllable systematization, of testing reinforced concrete based on ultrasonic guided-wave on rebar. In order to investigate how the deteriorated phenomena has the effects on the ultrasonic guided-wave propagating on the rebar, following experiments are conducted. Test pieces used for the experiments are made of bare steel rod and cylindrically pored mortar to be representing the actual reinforced concrete. Irradiating the end face of the rod with nanosecond pulsed laser makes the ultrasonic guided-wave induced, at the other end face, the guided wave signal is measured with ultrasonic receiver. One test piece is with no damage and the other is deteriorated test piece. The deterioration is made by electrolytic corroded method. The guided-wave signal from the deteriorated test piece is measured with respect to each energization time, the change in the waveform is investigated. Analyzing the results from the experiments above, it is found that the deterioration of rebar has remarkable effects on the guided-wave signal. The signal from test piece with no damage has strong peak at both specific frequency and lower region, on the other hand, signals from deteriorated test piece has only at the specific frequency depending on the diameter of the steel rod. Finally, discussion concerning with the experimental results and future perspective for remote controllable systematization of our method is carried out.

Study on crystalline rock aiming at evaluation method of long-term behavior of rock mass (Joint research)

Fukui, Katsunori*; Hashiba, Kimihiro*; Matsui, Hiroya

JAEA-Research 2017-010, 61 Pages, 2017/11

JAEA-Research-2017-010.pdf:16.86MB

JAEA has started this study as a collaboration study with Tokyo University from 2016. In the fiscal year of 2016, creep testing on Tage tuff was continuously conducted. Existing theory of rate process and stochastic process was modified to be applied to evaluate effects of water, and then the modified theory was validated based on the results of strength and creep tests performed under dry and wet conditions. Furthermore, effects of water contents on stress-strain curves were examined by uniaxial compression testing under various water content conditions.

Verification of alternative dew point hygrometer for CV-LRT in MONJU; Short- and long-term verification for capacitance-type dew point hygrometer (Translated document)

Ichikawa, Shoichi; Chiba, Yusuke; Ono, Fumiyasu; Hatori, Masakazu; Kobayashi, Takanori; Uekura, Ryoichi; Hashiri, Nobuo*; Inuzuka, Taisuke*; Kitano, Hiroshi*; Abe, Hisashi*

JAEA-Research 2017-001, 40 Pages, 2017/03

JAEA-Research-2017-001.pdf:5.19MB

In order to reduce the influence on a plant schedule of the MONJU by the maintenance of dew point hygrometers, The JAEA examined a capacitance type dew point hygrometer as an alternative dew point hygrometer for a lithium-chloride type dew point hygrometer which had been used at the CV-LRT in the MONJU. As a result of comparing a capacitance type dew point hygrometer with a lithium-chloride type dew point hygrometer at the CV-LRT (Atmosphere: nitrogen, Testing time: 24 hours), there weren't significant difference between a capacitance type dew point hygrometer and a lithium-chloride type dew point hygrometer. As a result of comparing a capacitance dew point hygrometer with a high-mirror-surface type dew point hygrometer for long term verification (Atmosphere: air, Testing time: 24 months), the JAEA confirmed that a capacitance type dew point hygrometer satisfied the instrument specification (2.04C) required by the JEAC4203-2008.

Visualization in response analyses for a nuclear power plant

Nakajima, Norihiro; Nishida, Akemi; Miyamura, Hiroko; Iigaki, Kazuhiko; Sawa, Kazuhiro

Kashika Joho Gakkai-Shi (USB Flash Drive), 36(Suppl.2), 4 Pages, 2016/10

Since nuclear power plants have dimensions approximately 100m and their structures are an assembly made up of over 10 million components, it is not convenient to experimentally analyze its behavior under strong loads of earthquakes, due to the complexity and hugeness of plants. The proposed system performs numerical simulations to evaluate the behaviors of an assembly like a nuclear facility. The paper discusses how to carry out visual analysis for assembly such as nuclear power plants. In a result discussion, a numerical experiment was carried out with a numerical model of High Temperature engineering Test Reactor of Japan Atomic Energy Agency and its result was compared with observed data. A good corresponding among them was obtained as a structural analysis of an assembly by using visualization. As a conclusion, a visual analytics methodology for assembly is discussed.

Progress of ITER full tungsten divertor technology qualification in Japan

Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Escourbiac, F.*; Hirai, Takeshi*; Kuznetsov, V.*

Fusion Engineering and Design, 98-99, p.1281 - 1284, 2015/10

https://doi.org/10.1016/j.fusengdes.2015.03.009

Japan Atomic Energy Agency (JAEA) is now devoting to development of Full-W ITER divertor outer vertical target (OVT), especially, PFU that needs to withstand the repetitive heat load as high as 20MW/m. JAEA have succeeded in demonstrating that the soundness of a bonding technology is sufficient for the full-W ITER divertor. For the development of bonding technology, the load carrying capability test on the W monoblock with a leg attachment to an OVT support structure was carried out and shows that the attachment can withstand against the uniaxial load more than 20 kN which is three times higher than the IO requirement. JAEA manufactured 6 small-scale mock-ups and tested under the repetitive heat load of 10 and 20 MW/m to examine the durability of the divertor structure including W tile bonding and the cooling tube. All of the mock-ups could survived 5000 cycles at 10 MW/m and 1000 cycles 20 MW/m with no failure such as debonding of the W tile and water leak from the cooling tube. The number of cycles at 20 MW/m is three times longer than the requirement of ITER divertor.

Structural analysis for assembly by integrating parts

Nakajima, Norihiro; Nishida, Akemi; Kawakami, Yoshiaki; Okada, Tatsuo*; Tsuruta, Osamu*; Sawa, Kazuhiro; Iigaki, Kazuhiko

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07

https://doi.org/10.1115/ICONE22-30251

Almost all industrial products are assembled from multiple parts. A nuclear facility is a large structure consisting of more than 10 million components. This paper discusses a method to analyze an assembly by gathering data on its component parts. Gathered data on component may identify ill conditioned meshes for connecting surfaces between components. These ill meshes are typified by nodal point disagreement in finite element discretization. A technique to resolve inconsistencies in data among the components is developed. By using this technique, structural analysis for an assembly can be carried out, and results can be obtained by the use of supercomputers, such as the K computer. Numerical results are discussed for components of the High Temperature Engineering Test Reactor.

Overview of design and R&D of test blankets in Japan

Enoeda, Mikio; Akiba, Masato; Tanaka, Satoru*; Shimizu, Akihiko*; Hasegawa, Akira*; Konishi, Satoshi*; Kimura, Akihiko*; Koyama, Akira*; Sagara, Akio*; Muroga, Takeo*

Fusion Engineering and Design, 81(1-7), p.415 - 424, 2006/02

https://doi.org/10.1016/j.fusengdes.2005.08.097

no abstracts in English

Plan and strategy for ITER blanket testing in Japan

Enoeda, Mikio; Akiba, Masato; Tanaka, Satoru*; Shimizu, Akihiko*; Hasegawa, Akira*; Konishi, Satoshi*; Kimura, Akihiko*; Koyama, Akira*; Sagara, Akio*; Muroga, Takeo*

Fusion Science and Technology, 47(4), p.1023 - 1030, 2005/05

https://doi.org/10.13182/FST05-A822

The Fusion Council of Japan has established the long-term research and development program of the blanket in 1999. In the program, the solid breeder blanket was selected as the primary candidate blanket of the fusion power demonstration plant in Japan. In the program, Japan Atomic Energy Research Institute (JAERI) has been nominated as a leading institute of the development of solid breeder blankets, in collaboration with universities, for the near term power demonstration plant, while, universities including National Institute for Fusion Science (NIFS) are assigned mainly to develop advanced blankets for longer term power plant development. In the long term research and development program, ITER blanket module testing is identified as the most important milestone, by which integrity of candidate blanket concepts and structures are evaluated. In Japan, universities, NIFS and JAERI cover a variety of types of blanket development. This paper presents a plan and strategy for ITER blanket module testing in Japan.