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JAEA Reports

Long term monitoring and evaluation of the excavation damaged zone induced around the wall of the shaft applying optical fiber sensor (Cooperative research)

Hata, Koji*; Niunoya, Sumio*; Uyama, Masao*; Nakaoka, Kenichi*; Fukaya, Masaaki*; Aoyagi, Kazuhei; Sakurai, Akitaka; Tanai, Kenji

JAEA-Research 2020-010, 142 Pages, 2020/11

JAEA-Research-2020-010.pdf:13.74MB
JAEA-Research-2020-010-appendix(DVD-ROM).zip:149.9MB

In the geological disposal study of high-level radioactive waste, it is suggested that the excavation damaged zone (EDZ) which is created around a tunnel by the excavation will be possible to be one of the critical path of radionuclides. Especially, the progress of cracks in and around the EDZ with time affects the safety assessment of geological disposal and it is important to understand the hydraulic change due to the progress of cracks in and around EDZ. In this collaborative research, monitoring tools made by Obayashi Corporation were installed at a total of 9 locations in the three boreholes near the depth of 370 m of East Shaft at the Horonobe Underground Research Laboratory constructed in the Neogene sedimentary rock. The monitoring tool consists of one set of "optical AE sensor" for measuring of the mechanical rock mass behavior and "optical pore water pressure sensor and optical temperature sensor" for measuring of groundwater behavior. This tool was made for the purpose of selecting and analyzing of AE signal waveforms due to rock fracture during and after excavation of the target deep shaft. As a result of analyzing various measurement data including AE signal waveforms, it is able to understand the information on short-term or long-term progress of cracks in and around EDZ during and after excavation in the deep shaft. In the future, it will be possible to carry out a study that contributes to the long-term stability evaluation of EDZ in sedimentary rocks in the deep part of the Horonobe Underground Research Laboratory by evaluation based on these analytical data.

JAEA Reports

Synthesis report on the R&D for the Horonobe Underground Research Laboratory; Project carried out during fiscal years 2015-2019

Nakayama, Masashi; Saiga, Atsushi; Kimura, Shun; Mochizuki, Akihito; Aoyagi, Kazuhei; Ono, Hirokazu; Miyakawa, Kazuya; Takeda, Masaki; Hayano, Akira; Matsuoka, Toshiyuki; et al.

JAEA-Research 2019-013, 276 Pages, 2020/03

JAEA-Research-2019-013.pdf:18.72MB

The Horonobe Underground Research Laboratory (URL) Project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of relevant disposal technologies for geological disposal of High-level Radioactive Waste through investigations of the deep geological environment within the host sedimentary rock at Horonobe Town in Hokkaido, north Japan. The investigations will be conducted in three phases, namely "Phase 1: Surface based investigations", "Phase 2: Construction phase" (investigations during construction of the underground facilities) and "Phase 3: Operation phase" (research in the underground facilities). According to the research plan described in the 3rd Mid- and Long- term Plan of JAEA, "Near-field performance study", "Demonstration of repository design option", and "Verification of crustal-movement buffering capacity of sedimentary rocks" are important issues of the Horonobe URL Project, and schedule of future research and backfill plans of the project will be decided by the end of 2019 Fiscal Year. The present report summarizes the research and development activities of these 3 important issues carried out during 3rd Medium to Long-term Research Phase.

Journal Articles

A Measurement method of long-term mechanical stability of support and rock mass after the excavation of galleries; Case study in Horonobe Underground Research Center

Aoyagi, Kazuhei; Sakurai, Akitaka; Miyara, Nobukatsu; Sugita, Yutaka; Tanai, Kenji

Shigen, Sozai Koenshu (Internet), 6(2), 7 Pages, 2019/09

no abstracts in English

Journal Articles

A Study on the hydro-mechanical behavior in the excavation damaged zone in shaft sinking at the Horonobe Underground Research Laboratory

Aoyagi, Kazuhei; Sakurai, Akitaka; Tanai, Kenji

Dai-46-Kai Gamban Rikigaku Ni Kansuru Shimpojiumu Koenshu (CD-ROM), p.142 - 147, 2019/01

This research presents the hydro-mechanical behavior of EDZ in shaft sinking in the Horonobe underground Research Laboratory on the basis of the results of in situ hydraulic tests, acoustic emission (AE) measurements, and hydro-mechanical coupling numerical analysis. The AE sources were distributed within 1.5 m into the shaft wall; and hydraulic conductivity in the EDZ is 2 to 4 orders of magnitudes higher than that in no fractured area. On the other hand, on the basis of the result of numerical analysis, the maximum extent of the EDZ is 1.5 m into the gallery wall. This result is almost consistent with the trend of acoustic emission measurement and hydraulic test.

Journal Articles

Development of residual thermal stress-relieving structure of CFC monoblock target for JT-60SA divertor

Tsuru, Daigo; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi

Fusion Engineering and Design, 98-99, p.1403 - 1406, 2015/10

 Times Cited Count:3 Percentile:26.04(Nuclear Science & Technology)

Journal Articles

Infrared thermography inspection for monoblock divertor target in JT-60SA

Nakamura, Shigetoshi; Sakurai, Shinji; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Sakasai, Akira; Tsuru, Daigo

Fusion Engineering and Design, 89(7-8), p.1024 - 1028, 2014/10

 Times Cited Count:4 Percentile:37.17(Nuclear Science & Technology)

Carbon Fiber Composite mono-block divertor target is required for power handling in JT-60SA. Heat removal capability of the target is degraded by joint defect which is induced in manufacturing process. For screening heat removal capability, infrared thermography inspection (IR inspection) is improved an accuracy for the target using threaded cooling tube. In IR inspection, the targets heated at 95$$^{circ}$$C by hot water in steady state condition are instantaneously cooled down by cold water flow of 5$$^{circ}$$C in three channels of test section. The heat removal capability of the targets is evaluated with comparing the transient thermal response time between defect-free and tested targets. A construction of a database for a correlation between the known defects, maximum surface temperatures in the heat load test and the IR inspection are successfully completed. Screening criteria is set with finite element methods based on the database.

Journal Articles

Manufacturing and development of JT-60SA vacuum vessel and divertor

Sakasai, Akira; Masaki, Kei; Shibama, Yusuke; Sakurai, Shinji; Hayashi, Takao; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Yokoyama, Kenji; Seki, Yohji; Shibanuma, Kiyoshi; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03

The JT-60SA vacuum vessel (VV) and divertor are key components for the performance requirements. Therefore the manufacturing and development of VV and divertor are in progress, inclusive of the superconducting magnets. The vacuum vessel has a double wall structure in high rigidity to withstand electromagnetic force at disruption and to keep high toroidal one-turn resistance. In addition, the double wall structure fulfills originally two functions. (1) The remarkable reduction of the nuclear heating in the superconducting magnets is made by boric-acid water circulated in the double wall. (2) The effective baking is enabled by nitrogen gas flow of 200$$^{circ}$$C in the double wall after draining of water. Three welding types were chosen for the manufacturing of the double wall structure VV to minimize deformation by welding. Divertor cassettes with fully water cooled plasma facing components were designed to realize the JT-60SA lower single null closed divertor. The divertor cassettes in the radio-active VV have been developed to ensure compatibility with remote handling (RH) maintenance in order to allow long pulse high performance discharges with high neutron yield. The manufacturing of divertor cassettes with typical accuracy of *1 mm has been successfully completed. Brazed CFC (carbon fiber composite) monoblock targets for a divertor target have been manufactured by precise control of tolerances inside CFC blocks. The infrared thermography test of monoblock targets has been developed as new acceptance inspection.

JAEA Reports

Establishment of reassembly technique of capsule type irradiation rig

Ichikawa, Shoichi; Abe, Kazuyuki; Haga, Hiroyuki; Kajima, Hisashi*; Sakurai, Satoshi*; Katsuyama, Kozo; Maeda, Koji; Nishinoiri, Kenji

JAEA-Technology 2011-032, 46 Pages, 2012/01

JAEA-Technology-2011-032.pdf:8.46MB

The assembly technique to the capsular irradiation rig newly developed was established. In the irradiation examination, the assembling disassembling and reassembling to PFB110 "B11(1), B11(2)" and PFB140 "B14" that built in Am-MOX fuel pin was achieved. The reassembly technique by recycling the irradiation material was established in the assembly of B11(2). This time, the assembly and disassembly of B11 (1) were reported. Moreover, the assembly of B14 which improved the assembly technology of B11 (1) was reported.

Journal Articles

Surface and interface studies by neutron reflectivity

Sakurai, Kenji*; Hino, Masahiro*; Takeda, Masayasu

Journal of the Vacuum Society of Japan, 53(12), p.747 - 752, 2010/12

Neutron reflectivity is a feasible probe for surfaces and interfaces. The technique has some common features to X-ray reflectivity, but at the same time it owns very unique and extremely attractive features, such as high-sensitivity to low Z elements and availability of magnetic structure analysis. The present article describes the recent activities of currently accessible neutron reflectometers in Japan.

JAEA Reports

Conceptual design of the SlimCS fusion DEMO reactor

Tobita, Kenji; Nishio, Satoshi*; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Uto, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; et al.

JAEA-Research 2010-019, 194 Pages, 2010/08

JAEA-Research-2010-019-01.pdf:48.47MB
JAEA-Research-2010-019-02.pdf:19.4MB

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m$$^{2}$$. This report covers various aspects of design study including systemic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept.

Journal Articles

Compact DEMO, SlimCS; Design progress and issues

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Kawashima, Hisato; Kurita, Genichi; Tanigawa, Hiroyasu; Nakamura, Hirofumi; Honda, Mitsuru; Saito, Ai*; Sato, Satoshi; et al.

Nuclear Fusion, 49(7), p.075029_1 - 075029_10, 2009/07

 Times Cited Count:135 Percentile:97.73(Physics, Fluids & Plasmas)

Recent design study on SlimCS focused mainly on the torus configuration including blanket, divertor, materials and maintenance scheme. For vertical stability of elongated plasma and high beta access, a sector-wide conducting shell is arranged in between replaceable and permanent blanket. The reactor adopts pressurized-water-cooled solid breeding blanket. Compared with the previous advanced concept with supercritical water, the design options satisfying tritium self-sufficiency are relatively scarce. Considered divertor technology and materials, an allowable heat load to the divertor plate should be 8 MW/m$$^{2}$$ or lower, which can be a critical constraint for determining a handling power of DEMO (a combination of alpha heating power and external input power for current drive).

Journal Articles

Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

Higashijima, Satoru; Sakurai, Shinji; Suzuki, Satoshi; Yokoyama, Kenji; Kashiwa, Yoshitoshi; Masaki, Kei; Shibama, Yusuke; Takechi, Manabu; Shibanuma, Kiyoshi; Sakasai, Akira; et al.

Fusion Engineering and Design, 84(2-6), p.949 - 952, 2009/06

 Times Cited Count:9 Percentile:53.48(Nuclear Science & Technology)

An upgrading device of JT-60 tokamak with fully superconducting coils (JT-60SA) is constructed under both the Japanese domestic program and the international program "Broader Approach". The maximum heat flux to JT-60SA divertor is estimated to 15 MW/m$$^{2}$$ for 100 s, and a monoblock-type CFC divertor armor is promising. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin OFHC-Cu buffer layer, and the brazed joints are essential for the armor. Metalization inside CFC monoblock is applied for further improvement, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace is also produced, and the half of the mock-ups could remove 15 MW/m$$^{2}$$ as required. This summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

Journal Articles

Simulation study for divertor design to handle huge exhaust power in the SlimCS DEMO reactor

Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Tobita, Kenji; Nishio, Satoshi; Sakurai, Shinji; Takenaga, Hidenobu

Nuclear Fusion, 49(6), p.065007_1 - 065007_7, 2009/06

 Times Cited Count:24 Percentile:67.65(Physics, Fluids & Plasmas)

By the SOLDOR/NEUT2D simulation for divertor design study on DEMO reactor, SlimCS, we estimated a prospect of handling huge exhausted power. Assuming the exhausted power 500 MW and ion out flux 0.5$$times$$10$$^{23}$$ s$$^{-1}$$ into the scrape-off-layer, the peak heat load is estimated to be 70 MW/m$$^{2}$$ on the outer target on the initial divertor design with introduction of moderate gas puff and Ar fraction. This value exceeds the allowable level 10 MW/m$$^{2}$$ being an initial design target. By installing the "${it V-shaped corner}$" in bottom of the outer divertor target, and using strong gas puffing or Ar impurity injection, the detached condition with high particle recycling and radiation loss conditions is formed, and the peak heat load is successfully reduced below 10 MW/m$$^{2}$$. It can also be demonstrated that the peak heat load is reduced exponentially with decrease of the exhaust power and reaches to 7 MW/m$$^{2}$$ at $$Q$$$$_{rm total}$$ = 300 MW for moderate gas puff flux and Ar fraction.

Journal Articles

SlimCS; Compact low aspect ratio DEMO reactor with reduced-size central solenoid

Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.

Nuclear Fusion, 47(8), p.892 - 899, 2007/08

 Times Cited Count:55 Percentile:86.05(Physics, Fluids & Plasmas)

The concept for a compact DEMO reactor named "SlimCS" is presented. Distinctive features of the concept is low aspect ratio ($$A$$ = 2.6) and use of a reduced-size center solenoid (CS) which has a function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field (TF) coil system which contributes to reducing the weight and construction cost of the reactor. SlimCS is as compact as advanced commercial reactor designs such as ARIES-RS and produces 1 GWe in spite of moderate requirements for plasma parameters. Merits of low-$$A$$, i.e. vertical stability for high elongation and high beta limit are responsible for such reasonable physics requirements.

Journal Articles

Concept of compact low aspect ratio Demo reactor, SlimCS

Tobita, Kenji; Nishio, Satoshi; Sato, Masayasu; Sakurai, Shinji; Hayashi, Takao; Shibama, Yusuke; Isono, Takaaki; Enoeda, Mikio; Nakamura, Hirofumi; Sato, Satoshi; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2006/10

no abstracts in English

Journal Articles

Case study on tritium inventory in the fusion DEMO plant at JAERI

Nakamura, Hirofumi; Sakurai, Shinji; Suzuki, Satoshi; Hayashi, Takumi; Enoeda, Mikio; Tobita, Kenji; DEMO Plant Design Team

Fusion Engineering and Design, 81(8-14), p.1339 - 1345, 2006/02

 Times Cited Count:50 Percentile:94.68(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Concept of core and divertor plasma for fusion DEMO plant at JAERI

Sato, Masayasu; Sakurai, Shinji; Nishio, Satoshi; Tobita, Kenji; Inoue, Takashi; Nakamura, Yukiharu; Shinya, Kichiro*; Fujieda, Hirobumi*; DEMO Plant Design Team

Fusion Engineering and Design, 81(8-14), p.1277 - 1284, 2006/02

 Times Cited Count:14 Percentile:68.21(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Design study of fusion DEMO plant at JAERI

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.

Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02

 Times Cited Count:117 Percentile:99.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Divertor power handling in a low aspect ratio tokamak reactor

Sakurai, Shinji; Tobita, Kenji; Nishio, Satoshi

Purazuma, Kaku Yugo Gakkai-Shi, 80(11), p.955 - 958, 2004/11

no abstracts in English

Journal Articles

Advanced fusion technologies developed for JT-60 superconducting Tokamak

Sakasai, Akira; Ishida, Shinichi; Matsukawa, Makoto; Akino, Noboru; Ando, Toshinari*; Arai, Takashi; Ezato, Koichiro; Hamada, Kazuya; Ichige, Hisashi; Isono, Takaaki; et al.

Nuclear Fusion, 44(2), p.329 - 334, 2004/02

no abstracts in English

42 (Records 1-20 displayed on this page)