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Journal Articles

Evolution of the reaction and alteration of granite with Ordinary Portland cement leachates; Sequential flow experiments and reactive transport modelling

Bateman, K.*; Murayama, Shota*; Hanamachi, Yuji*; Wilson, J.*; Seta, Takamasa*; Amano, Yuki; Kubota, Mitsuru*; Ouchi, Yuji*; Tachi, Yukio

Minerals (Internet), 12(7), p.883_1 - 883_20, 2022/07

Journal Articles

Temporal variability of $$^{137}$$Cs concentrations in coastal sediments off Fukushima

Suzuki, Shotaro*; Amano, Yosuke*; Enomoto, Masahiro*; Matsumoto, Akira*; Morioka, Yoshiaki*; Sakuma, Kazuyuki; Tsuruta, Tadahiko; Kaeriyama, Hideki*; Miura, Hikaru*; Tsumune, Daisuke*; et al.

Science of the Total Environment, 831, p.154670_1 - 154670_15, 2022/07

 Times Cited Count:0 Percentile:0(Environmental Sciences)

Journal Articles

Evolution of the reaction and alteration of mudstone with ordinary Portland cement leachates; Sequential flow experiments and reactive-transport modelling

Bateman, K.; Murayama, Shota*; Hanamachi, Yuji*; Wilson, J.*; Seta, Takamasa*; Amano, Yuki; Kubota, Mitsuru*; Ouchi, Yuji*; Tachi, Yukio

Minerals (Internet), 11(9), p.1026_1 - 1026_23, 2021/09

 Times Cited Count:1 Percentile:42.19(Geochemistry & Geophysics)

Journal Articles

Holding force tests of Curie Point Electro-Magnet in hot gas for passive shutdown system

Matsunaga, Shoko*; Matsubara, Shinichiro*; Kato, Atsushi; Yamano, Hidemasa; D$"o$derlein, C.*; Guillemin, E.*; Hirn, J.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05

This paper presents a design of Curie Point Electro-Magnet (CPEM) which will be installed as a passive shutdown system for a French Sodium-cooled Fast Reactor (ASTRID) development program which is conducted in collaboration between France and Japan. To confirm CPEM design validity, a qualification program for CPEM is developed on the basis of past comprehensive test series of Self-Actuated Shutdown System (SASS) in Japan. The main outcome of this paper is results of holding force tests in hot gas, which satisfy design requirements. Moreover, the result of a numerical magnetic field analysis showed the same tendency as that of the holding force test.

Journal Articles

Determination of fusion barrier distributions from quasielastic scattering cross sections towards superheavy nuclei synthesis

Tanaka, Taiki*; Narikiyo, Yoshihiro*; Morita, Kosuke*; Fujita, Kunihiro*; Kaji, Daiya*; Morimoto, Koji*; Yamaki, Sayaka*; Wakabayashi, Yasuo*; Tanaka, Kengo*; Takeyama, Mirei*; et al.

Journal of the Physical Society of Japan, 87(1), p.014201_1 - 014201_9, 2018/01

 Times Cited Count:10 Percentile:67.6(Physics, Multidisciplinary)

Excitation functions of quasielastic scattering cross sections for the $$^{48}$$Ca + $$^{208}$$Pb, $$^{50}$$Ti + $$^{208}$$Pb, and $$^{48}$$Ca + $$^{248}$$Cm reactions were successfully measured by using the gas-filled recoil-ion separator GARIS. Fusion barrier distributions were extracted from these data, and compared with the coupled-channels calculations. It was found that the peak energies of the barrier distributions for the $$^{48}$$Ca + $$^{208}$$Pb and $$^{50}$$Ti + $$^{208}$$Pb systems coincide with those of the 2n evaporation channel cross sections for the systems, while that of the $$^{48}$$Ca + $$^{248}$$Cm is located slightly below the 4n evaporation ones. This results provide us helpful information to predict the optimum beam energy to synthesize superheavy nuclei.

Journal Articles

Safety evaluation of self actuated shutdown system for Gen-IV SFR

Saito, Hiroyuki*; Yamada, Yumi*; Oyama, Kazuhiro*; Matsunaga, Shoko*; Yamano, Hidemasa; Kubo, Shigenobu

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

A self-actuated shutdown system (SASS) is a passive device, which can detach a control rod for reactor shutdown in response to excessive increase in coolant temperature. Since a detachment temperature, which triggers release of a control rod, and a response time are identified as important parameters for validity analyses, this study focused on investigation of the response time and the detachment temperature, and safety analysis to see feasibility of the SASS in low power. For this purpose, design modifications were made to shorten the response time and three-dimensional thermal-hydraulic analysis in a low power operation was carried out in order to confirm the response time. The resulting detachment temperature level is lower than previous studies, leading to improved safety parameters. Based on improved parameters, a safety analysis to see feasibility of the SASS in low power was carried out. From this safety evaluation, it was confirmed that core damage can be prevented by the SASS with flow collector in the case of LOF type ATWS event.

JAEA Reports

Proceedings of the 2013 Symposium on Nuclear Data; November 14-15, 2013, Research Institute of Nuclear Engineering University of Fukui, Tsuruga, Fukui, Japan

Yamano, Naoki*; Iwamoto, Osamu; Nakamura, Shoji; Kunieda, Satoshi; Van Rooijen, W.*; Koura, Hiroyuki

JAEA-Conf 2014-002, 209 Pages, 2015/02

JAEA-Conf-2014-002.pdf:64.24MB

The 2013 Symposium on Nuclear Data was held at Research Institute of Nuclear Engineering, University of Fukui, on 14th and 15th of November 2013. The Nuclear Data Division of the Atomic Energy Society of Japan and Research Institute of Nuclear Engineering, University of Fukui organize this symposium in cooperation with Nuclear Science and Engineering Center of Japan Atomic Energy Agency and the Chubu Branch of Atomic Energy Society of Japan. In the oral sessions, papers were presented on topics of progress in neutron cross-section measurement and analysis, application of nuclear data, recent topics on nuclear data measurement and theory, and progress in studies of high-energy nuclear reactions. In the poster session, papers were presented concerning experiments, evaluations, benchmark tests and applications. This report consists of total 35 papers including 14 oral presentations and 21 poster presentations.

Journal Articles

An Investigation on debris bed self-leveling behavior with non-spherical particles

Cheng, S.; Tagami, Hirotaka; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Takeda, Shohei*; Nishi, Shimpei*; Nishikido, Tatsuya*; Zhang, B.*; Matsumoto, Tatsuya*; et al.

Journal of Nuclear Science and Technology, 51(9), p.1096 - 1106, 2014/09

AA2013-0303.pdf:1.68MB

 Times Cited Count:18 Percentile:82.78(Nuclear Science & Technology)

Journal Articles

Experimental study and empirical model development for self-leveling behavior of debris bed using gas-injection

Cheng, S.; Tagami, Hirotaka; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Nakamura, Yuya*; Takeda, Shohei*; Nishi, Shimpei*; Zhang, B.*; Matsumoto, Tatsuya*; et al.

Mechanical Engineering Journal (Internet), 1(4), p.TEP0022_1 - TEP0022_16, 2014/08

JAEA Reports

The Investigation related to the study on the method to withdraw the in-vessel transfer machine; Observation of the structure in the reactor vessel of the fast breeder reactor Monju

Harigae, Hitoshi; Takagi, Tsuyohiko; Hamano, Tomoharu; Nakamura, Shoichi; Oba, Toshio; Ebashi, Masaaki; Okuda, Eiichi; Kinoshita, Tomonobu

JAEA-Technology 2013-014, 150 Pages, 2013/07

JAEA-Technology-2013-014.pdf:24.38MB

In-Vessel Transfer Machine (IVTM) came off from the gripper claw in the Auxiliary Handling Machine (AHM) and fell at a height of approximately two meters during a withdrawal work of the IVTM in the Fast Breeder Reactor (FBR) Monju. The withdrawal work of IVTM from the reactor vessel by AHM was performed. The work, however, was suspended due to the excessive load alarm. To grasp the situation of the IVTM fall, observation of the machine was necessary. An interior observation and an exterior observation of the dropped IVTM were performed. As a result of these observations, the radially deformed lower end of the upper guide tube was observed at the connection part, and it was jammed in the fuel throat sleeve when the dropped IVTM was withdrawn. Based on this information, the IVTM could be safely withdrawn from the reactor vessel with the fuel throat sleeve.

Journal Articles

Recent knowledge from an experimental investigation on self-leveling behavior of debris bed

Cheng, S.; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Nakamura, Yuya*; Takeda, Shohei*; Nishi, Shimpei*; Zhang, B.*; Matsumoto, Tatsuya*; Morita, Koji*

Proceedings of 21st International Conference on Nuclear Engineering (ICONE-21) (DVD-ROM), 8 Pages, 2013/07

Journal Articles

Production of mutants by ion beam irradiation in ${it Dahlia}$ spp.

Uyama, Yoshihide*; Oya, Hirotaka*; Amano, Yoshinori*; Kashimoto, Koichi*; Hatano, Shoji*; Nozawa, Shigeki; Yoshihara, Ryohei*; Hase, Yoshihiro; Narumi, Issei

JAEA-Review 2012-046, JAEA Takasaki Annual Report 2011, P. 102, 2013/01

Journal Articles

LDV flow measurement of a deflected inflow using a 1/10-scale hot-log piping test facility of a primary circuit hot-leg piping in a sodium-cooled fast reactor

Iwamoto, Yukiharu*; Kondo, Manabu*; Ogawa, Shota*; Tanaka, Masaaki; Yamano, Hidemasa

Nihon Kikai Gakkai Rombunshu, B, 78(792), p.1383 - 1387, 2012/08

LDV measurements in a 90 degrees elbow which curvature radius coincides with the diameter have been conducted. This paper especially focuses on a result of the deflected inflow, comparing with a result of the short pipe. The result shows that the deflected inflow reinforced a convex velocity distribution occurring near the curvature inside in the downstream region, concluding that the deflected inflow promotes the secondary flow of Prandtl's first kind in the elbow. Its Strouhal number increases to 0.6 from 0.5, compared with the short pipe case. Results of frequency analyses are also shown for other cases that we have been examined. Dominant Strouhal numbers in most of the cases become 0.5, except for 0.6 in cases of the inflow from the long pipe and deflector. This frequency shift might be related with the boundary layer size and the local flow velocity, since the corresponding fluctuation is caused by vortex shedding from the boundary layer at the elbow inside.

Journal Articles

Safety design approach for a large-scale Japan Sodium-cooled Fast Reactor (JSFR)

Kotake, Shoji*; Yamano, Hidemasa; Sagayama, Yutaka

Fusion Science and Technology, 61(1T), p.137 - 143, 2012/01

 Times Cited Count:1 Percentile:11.11(Nuclear Science & Technology)

The present paper describes safety goals and principles for Generation IV energy systems, with emphasis on prevention and mitigation against severe accidents in the safety design corresponding to Level 4 of the defense-indepth architecture. Consistent with them, a deterministic safety design approach has been applied to the Japan Sodium-cooled Fast Reactor (JSFR) with the complimentary use of probabilistic approach. The JSFR safety design principle has also been developed with safety design features corresponding to essential safety functions, such as reactor shutdown, decay heat removal and containment. Design principle against chemical activity of sodium is also discussed both on the isolation from the reactor core safety and the contribution to the plant reliability.

Journal Articles

Safety design requirements for safety systems and components of JSFR

Kubo, Shigenobu*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kotake, Shoji

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 12 Pages, 2012/00

In order to embody a safety design, a higher level safety principle was broken down into a set of design requirements for each safety related system, structure and component (SSC). This paper will present an output of the safety requirements for safety related SSCs of JSFR.

Journal Articles

Fault zone determination and bedrock classification through multivariate analysis; Case study using a dataset from a deep borehole in the Mizunami Underground Research Laboratory

Abumi, Kensho*; Amano, Kenji; Koike, Katsuaki*; Tsuruta, Tadahiko; Matsuoka, Toshiyuki

Joho Chishitsu, 22(4), p.171 - 188, 2011/12

Fault zones are treated as essential elements for evaluating the underground geological environment and the engineering performance of rocks. Because of the limitations to borehole investigations, it is not always possible to obtain sufficient, high-quality geological data. In addition, the evaluation of results may differ depending on various factors such as geological conditions and skill of the engineer. Such uncertainty can lead to difficulty in evaluation and understanding of the geological environment at depths and in the decision-making and planning of underground construction, which, as a result, may increases potential risks during construction. To reduce the uncertainty, this study proposes a correct selection method of data item for multivariable analyses composed of principal component analysis and clustering method using a deep borehole data. Utilizing this method and the analyses, the rocks could be accurately classified depending upon their geological characteristics.

Journal Articles

Safety design requirements for safety systems and components of JSFR

Kubo, Shigenobu*; Shimakawa, Yoshio*; Yamano, Hidemasa; Kotake, Shoji

Journal of Nuclear Science and Technology, 48(4), p.547 - 555, 2011/04

Safety design requirements for JSFR were summarized taking the development targets of FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF and basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global-standard. The development targets for safety and reliability are set based on that of FaCT. Namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth philosophy is used as the basic safety design principle. General features of the safety design requirements are (1) Achievement of higher reliability, (2) Achievement of higher inspectability and maintainability, (3) Introduction of passive safety features, (4) Reduction of operator action needs, (5) Design consideration against Beyond Design Basis Events, (6) In Vessel Retention of degraded core materials, (7) Prevention and mitigation against sodium chemical reactions, (8) Design against external events. Current specific requirements for the each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop type large output power plant with mixed oxide fuelled core.

Journal Articles

Several-MeV $$gamma$$-ray generation at NewSUBARU by laser Compton backscattering

Amano, Sho*; Horikawa, Ken*; Ishihara, Kazuki*; Miyamoto, Shuji*; Hayakawa, Takehito; Shizuma, Toshiyuki; Mochizuki, Takayasu*

Nuclear Instruments and Methods in Physics Research A, 602(2), p.337 - 341, 2009/04

 Times Cited Count:61 Percentile:97.31(Instruments & Instrumentation)

A new laser Compton $$gamma$$ ray source in the energy range of several-MeV was developed using a CO$$_2$$ laser at the NewSUBARU electron storage ring. When the electron energies were 974, 1220 and 1460 MeV, the maximum $$gamma$$ ray energies were measured to be 1.72, 2.72, and 3.91 MeV. The luminosity of the $$gamma$$ rays was 7300 photon/mA/W/s and a flux of 5.8$$times$$10$$^6$$ photon/s was achieved. These performances agreed with calculations. This generation of $$gamma$$ rays is in a no-loss mode for the storage electrons, and the maximum flux was limited only by the power of the laser.

Journal Articles

Study on flow-induced-vibration evaluation of large-diameter pipings in a sodium-cooled fast reactor, 2; A Large-Eddy Simulation of turbulent flow in a short-elbow pipe

Eguchi, Yuzuru*; Murakami, Takahiro*; Ohshima, Hiroyuki; Yamano, Hidemasa; Kotake, Shoji

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 8 Pages, 2008/11

The unsteady turbulent flow in a short-elbow pipe to be employed in a Japanese sodium-cooled fast breeder reactor was computed to examine the fundamental features of the flow, especially, pressure fluctuation to cause unsteady fluid force on the pipe. An FEM-based large-eddy simulation code, named SMART-fem, was used for the computation. The results at Re=3.2$$times$$10$$^{5}$$ and 1.2$$times$$10$$^{6}$$ show that two separation regions exist on the inner urvature of the elbow around 45-degree (middle of elbow) and 90-degree (end of elbow) positions. The statistical quantities of pressure fluctuation such as deviation, skewness and flatness were computed and analyzed, showing that there exist two symmetric regions of significant pressure fluctuation on the wall of inner curvature of the elbow. It has turned out that the pressure loss coefficient of the elbow pipe agrees well among the computation, experiment and authoritative reference data.

Journal Articles

Study on flow-induced-vibration evaluation of large-diameter pipings in a sodium-cooled fast reactor, 4; Experiments on the 1/10-scale hot leg test facility in Reynolds number of 50000 and 320000

Iwamoto, Yukiharu*; Yasuda, Kazunori*; Sogo, Motosuke*; Yamano, Hidemasa; Kotake, Shoji

Proceedings of 6th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-6) (USB Flash Drive), 6 Pages, 2008/11

Pressure measurement, laser Doppler velocimetry (LDV) and flow visualization were carried out using the 1/10-scale model of a hot leg piping installed in a Japanese sodium-cooled fast breeder reactor. LDV measurement with Reynolds number of 50000 showed the following results: (1) A flow separation was confirmed in the region between 45 degrees from the elbow inlet and 0.3 times of pipe diameter downstream of the elbow. (2) There appeared two kinds of fluctuations in the present study. In the case of Reynolds number of 320000, it was found that the height of the flow separation downstream of the elbow became smaller, since the inertia of the flow became superior to the inverse pressure gradient.

68 (Records 1-20 displayed on this page)