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Journal Articles

Study on the seismic buckling evaluation method for Mod. 9Cr-1Mo steel cylindrical vessel

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Kubo, Koji*; Sato, Kenichiro*; Wakai, Takashi; Shimomura, Kenta

Nippon Kikai Gakkai M&M 2017 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.591 - 595, 2017/10

no abstracts in English

Journal Articles

Flow-induced vibration evaluation of primary hot-leg piping in advanced loop-type sodium-cooled fast reactor for demonstration

Yamano, Hidemasa; Xu, Y.*; Sago, Hiromi*; Hirota, Kazuo*; Baba, Takeo*

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1029 - 1038, 2016/04

This study conducted the flow-induced vibration evaluation of the primary hot-leg piping in the demonstration reactor design of advanced loop-type sodium-cooled fast reactor in order to confirm the integrity of the piping. Following the description of the primary hot-leg piping design and a design guideline of the flow-induced vibration evaluation, this paper describes mainly the flow-induced vibration evaluation and thereby the integrity assessment. In the fatigue evaluation for the flow-induced vibration, the pipe stresses considering the stress concentration factor and so on, at representative locations were less than the design fatigue limit. Therefore, this evaluation confirmed the integrity of the primary hot-leg piping in the demonstration reactor.

Journal Articles

Development of proposed guideline of flow-induced vibration evaluation for hot-leg piping in a sodium-cooled fast reactor

Sakai, Takaaki; Yamano, Hidemasa; Tanaka, Masaaki; Ono, Ayako; Ohshima, Hiroyuki; Kaneko, Tetsuya*; Hirota, Kazuo*; Sago, Hiromi*; Xu, Y.*; Iwamoto, Yukiharu*; et al.

Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 13 Pages, 2013/05

The development of flow-induced vibration evaluation methodology has reached a milestone that separate-effect experimental data under a high Reynolds number regime including swirl and deflected inflow conditions are available for the validation of the methodology. On the other hand, technical standards are desirable to be documented for designers of sodium-cooled fast reactors. From such a background, the documentation of a flow-induced vibration design guideline has been made for the hot-leg piping of Japan sodium-cooled fast reactor. This paper describes the design guideline of the flow-induced vibration evaluation methodology, which has been informed from main separate-effect experiments, as well as supplemental interpretation for the guideline.

Journal Articles

Experimental study for the proposal of design measures against cover gas entrainment and vortex cavitation with 1/11th scale reactor upper sodium plenum model of Japan Sodium-cooled Fast Reactor

Yoshida, Kazuhiro*; Sakata, Hideyuki*; Sago, Hiromi*; Shiraishi, Tadashi*; Oyama, Kazuhiro*; Hagiwara, Hiroyuki*; Yamano, Hidemasa; Yamamoto, Tomohiko

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12

To prevent the vortex cavitations, asymmetric flow in the upper plenum due to the radial slit with upper internal structure (UIS) has been mitigated by installing a cylindrical structure named as dummy plug instead of the fuel handling machine only used for refueling period. In this study, the extended brim and the division plate at the slit of UIS have been proposed in order to improve flow pattern in upper plenum for the purpose of the vortex cavitation prevention.

Journal Articles

Effect of swirl inflow on flow pattern and pressure fluctuation onto a single-elbow pipe in Japan Sodium-cooled Fast Reactor

Yamano, Hidemasa; Sago, Hiromi*; Hirota, Kazuo*; Hayakawa, Satoshi*; Xu, Y.*; Tanaka, Masaaki; Sakai, Takaaki

Journal of Fluid Science and Technology (Internet), 7(3), p.329 - 344, 2012/09

As part of the development of a flow-induced vibration evaluation methodology for the primary cooling piping in Japan Sodium-cooled Fast Reactor, important factors were discussed in evaluating the flow-induced vibration for the hot-leg piping. To investigate a complex flow near the inlet of the hot-leg piping, a reactor scale numerical analysis was carried out for the reactor upper plenum flow, which was simulated in a 1/10-scale reactor upper plenum experiment. Based on this analysis, experimental conditions on swirl inflow and deflected inflow that were identified as important factors were determined for flow-induced vibration experiments simulating only the hot-leg piping. In this study, the effect of the swirl inflow on flow pattern and pressure fluctuation onto the pipe wall was investigated in a 1/3-scale hot-leg pipe experiment. The experiment has indicated less significant for the pressure fluctuations, while the flow separation region was slightly influenced by the swirl flow. Computational fluid dynamics simulation results also appear in this paper, focusing on its applicability to the hot-leg piping experiments.

Journal Articles

Flow pattern and pressure fluctuation characteristics on the 1/3 scale hot-leg piping experiments of a primary circuit hot-leg piping in a sodium-cooled fast reactor

Sago, Hiromi*; Shiraishi, Tadashi*; Watakabe, Hisato*; Xu, Y.*; Aizawa, Kosuke; Yamano, Hidemasa

Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1378 - 1382, 2012/08

A conceptual design study of the Japan Sodium-cooled Fast Reactor (JSFR) is in progress in "the Fast Reactor Cycle Technology Development (FaCT) project", and a two-loop primary system is adopted in order to economize the plant construction cost. In the JSFR the pipe thickness is designed to be considerably thinner and the mean sodium velocity increases. To understand the behavior of flow-induced vibration that is derived from the hydraulic characteristics under high Reynolds number conditions experiments were performed to evaluate and confirm the integrity.

Journal Articles

Prediction of unsteady flow field in a primary circuit hot-leg piping of a sodium-cooled fast reactor

Tanaka, Masaaki; Sago, Hiromi*; Iwamoto, Yukiharu*; Ebara, Shinji*; Ono, Ayako; Murakami, Takahiro*; Hayakawa, Satoshi*

Nippon Kikai Gakkai Rombunshu, B, 78(792), p.1392 - 1396, 2012/08

A study on flow induced vibration in the primary cooling system of Japan Sodium cooled Fast Reactor (JSFR) consisting of large diameter pipe and pipe elbow with short curvature radius ("short-elbow") has been conducted. Flow-induced vibration in the short-elbow is an important issue in design study of JSFR, because it may affect to structural integrity of the pipe. In this paper, unsteady flow characteristics in the JSFR short-elbow pipe related to the large-scale eddy motion were estimated based on knowledge from existing studies for curved pipes and scaled water experiments and numerical simulations for the JSFR hot-leg piping.

Journal Articles

Unsteady elbow pipe flow to develop a flow-induced vibration evaluation methodology for JSFR

Yamano, Hidemasa; Tanaka, Masaaki; Ono, Ayako; Murakami, Takahiro*; Iwamoto, Yukiharu*; Yuki, Kazuhisa*; Sago, Hiromi*; Hayakawa, Satoshi*

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

This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in JSFR, in particular emphasizing on recent R&D activities that investigate unsteady elbow pipe flow. The experiment using the 1/3-scale test section was performed to investigate the effect of swirl flow at the inlet. Although the flow separation region was distorted at the downstream from the elbow, the experiment clarified that the effect of swirl flow on pressure fluctuation onto the pipe wall was not significant. The simulation revealed that Reynolds number scarcely affects flow patterns and flow velocity distributions.

Journal Articles

Study on flow induced vibration evaluation for a large scale JSFR piping, 2; Vibration analysis in 1/3 scale hot-leg piping experiments under swirl inflow conditions

Baba, Takeo*; Hirota, Kazuo*; Sago, Hiromi*; Yamano, Hidemasa; Aizawa, Kosuke; Xu, Y.*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 9 Pages, 2011/05

Journal Articles

Study on flow induced vibration evaluation for a large scale JSFR piping, 3; Pressure fluctuation characteristics in 1/3 scale hot-leg piping experiments under deflected inflow conditions due to UIS structures

Sago, Hiromi*; Shiraishi, Tadashi*; Watakabe, Hisato*; Xu, Y.*; Aizawa, Kosuke; Yamano, Hidemasa

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/05

Journal Articles

Study on flow induced vibration evaluation for a large scale JSFR piping, 1; Current status of flow induced vibration evaluation for hot-leg piping

Yamano, Hidemasa; Sakai, Takaaki; Tanaka, Masaaki; Sago, Hiromi*; Hirota, Kazuo*; Xu, Y.*

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 8 Pages, 2011/05

Journal Articles

Unsteady elbow pipe flow to develop a flow-induced vibration evaluation methodology for Japan sodium-cooled fast reactor

Yamano, Hidemasa; Tanaka, Masaaki; Murakami, Takahiro*; Iwamoto, Yukiharu*; Yuki, Kazuhisa*; Sago, Hiromi*; Hayakawa, Satoshi*

Journal of Nuclear Science and Technology, 48(4), p.677 - 687, 2011/04

This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in Japan Sodium-cooled Fast Reactor (JSFR), in particular emphasizing on recent R&D activities that investigate unsteady elbow pipe flow.

Journal Articles

Technological feasibility of two-loop cooling system in JSFR

Yamano, Hidemasa; Kubo, Shigenobu*; Kurisaka, Kenichi; Shimakawa, Yoshio*; Sago, Hiromi*

Nuclear Technology, 170(1), p.159 - 169, 2010/04

 Times Cited Count:15 Percentile:22.85(Nuclear Science & Technology)

An advanced large-scale sodium-cooled fast reactor (named JSFR) adopts an innovative two-loop cooling system. This cooling system design gives rise to major technological issues: hydraulic and structural integrity due to the increase in one-loop coolant flow rate, safety design against the break or failure in one-loop piping and ensuring the reliability of decay heat removal system. The present paper describes that the piping structural integrity due to flow-induced vibration has been investigated using a 1/3-scale hot-leg piping test. The structural integrity of the hot-leg piping in the JSFR design has been confirmed by a flow-induced-vibration analytical methodology, verified with the experimental data. Additional experimental results have revealed that hydraulic issues including gas entrainment and vortex cavitation could be prevented by some design measures. By applying appropriate safety design, the two-loop system has been confirmed to be valid against the break or failure in one-loop piping by a safety evaluation in this study. The decay heat removal system with natural circulation is designed in conformity with the two-loop system by introducing adequate safety designs. In this paper, the validity of this decay heat removal system is given by a probabilistic safety assessment and safety evaluation.

Journal Articles

Pressure fluctuation characteristics of the short-radius elbow pipe for FBR in the postcritical Reynolds regime

Shiraishi, Tadashi*; Watakabe, Hisato*; Sago, Hiromi*; Yamano, Hidemasa

Journal of Fluid Science and Technology (Internet), 4(2), p.430 - 441, 2009/00

For the Japan Sodium-cooled Fast Reactor, an experimental study on the fluctuating pressure of the hot legs was carried out with tests using a 1/3-scale model. The total resistance coefficient is consistent with the published data, and our research has given some additional data up to the Reynolds number of 8.0$$times$$10$$^{6}$$. The flow pattern in the postcritical regime is independent of a Reynolds number. The statistical examination revealed that fluctuating pressures on the pipe wall depend on the mean velocity but not on the viscosity of the fluid. Negative spikes of pressure appeared for high velocity. Based on these experimental data it is concluded that, there are similarity laws for the scale of a model and the property of fluid, but not for the velocity in the pipe. Theoretical considerations also gave a discussion how to extrapolate the fluctuating pressure to the actual hot-leg conditions.

Journal Articles

Study on flow-induced-vibration evaluation of the large-diameter pipings in a sodium-cooled fast reactor, 3; Random vibration analysis method based on turbulence energy calculated by CFD

Hirota, Kazuo*; Ishitani, Yoshihide*; Nishida, Keigo*; Sago, Hiromi*; Xu, Y.*; Yamano, Hidemasa; Nakanishi, Shigeyuki; Kotake, Shoji

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

CFD simulation using the Reynolds stress model was performed to evaluate turbulence-induced forces on the piping. The turbulence energy with the CFD simulation was compared with pressure fluctuation distributions obtained by the test with a 1/3 scale elbow simulating the JSFR hot-leg piping. The profile of turbulence energy was good agreement with that of the pressure fluctuation. The magnitude of pressure fluctuation can also be estimated from calculated turbulence energy multiplied by a certain coefficient. In the vibration analysis, the power spectrum density (PSD) of the pressure fluctuation was derived from the measured normalized PSD multiplied by the coefficient. The vibration analysis method was proposed based on the PSDs derived by the above procedure and correlation lengths. The analysis results of vibration response showed good agreement with the flow-induced-vibration test results, thereby it can be said that the vibration analysis method developed in this study is valid.

Journal Articles

Pressure fluctuation characteristics of the short-radius elbow pipe for FBR in the postcritical Reynolds regime

Shiraishi, Tadashi*; Watakabe, Hisato*; Sago, Hiromi*; Kotake, Shoji; Yamano, Hidemasa

Proceedings of 2nd International Conference on Jets, Wakes and Separated Flows (ICJWSF 2008) (CD-ROM), 11 Pages, 2008/09

For the JAEA sodium-cooled fast reactor, an experimental study on the fluctuating pressure of the hot legs was carried out with tests using a 1/3-scale model. The total resistance coefficient is consistent with the published data, and our research has given some additional data up to the Reynolds number of 8.0$$times$$10$$^6$$. The flow pattern in the postcritical regime is independent of a Reynolds number. The statistical examination revealed that fluctuating pressures on the pipe wall depend on the mean velocity but not on the viscosity of the fluid. Negative spikes of pressure appeared for high velocity. Based on these experimental data it is revealed that, there are similarity laws for the scale of a model and the property of fluid, but not for the velocity in the pipe. We also discussed how to extrapolate the fluctuating pressure to the actual hot-leg conditions.

Journal Articles

Development of advanced loop-type fast reactor in Japan, 2; Technological feasibility of two-loop cooling system in JSFR

Yamano, Hidemasa; Kubo, Shigenobu; Kurisaka, Kenichi; Shimakawa, Yoshio*; Sago, Hiromi*

Proceedings of 2008 International Congress on Advances in Nuclear Power Plants (ICAPP '08) (CD-ROM), p.469 - 504, 2008/06

The conceptual design of an advanced sodium-cooled fast reactor (named JSFR) is currently carried out by the Japan Atomic Energy Agency (JAEA). In general, a large-scale nuclear reactor (approximately 1.5 GWe class) tended to increase in the number of loops (e.g., four loops in Super Ph$'e$nix and APWR), while the JSFR adopts a two-loop cooling system that enables significantly reducing a plant construction cost resulting from decreasing in material amount of the nuclear steam supply system and in the reactor building volume. This paper describes technological feasibility of the two-loop cooling system in JSFR; especially, focused on flow-induced vibration of piping, safety analysis and decay heat removal system.

Journal Articles

Flow-induced vibration of a large-diameter elbow piping in high Reynolds number range; Random force measurement and vibration analysis

Hirota, Kazuo*; Ishitani, Yoshihide*; Nakamura, Tomomichi*; Shiraishi, Tadashi*; Sago, Hiromi*; Yamano, Hidemasa; Kotake, Shoji

Proceedings of 9th International Conference on Flow-induced Vibrations (FIV 2008), 6 Pages, 2008/00

The present study is intended to grasp flow-induced vibration characteristics in the piping by newly taken experimental data as well as to verify a vibration analysis tool with its data. It was found that a flow velocity-dependent periodic phenomenon with maximum random vibration force appears in the downstream region of the elbow. The Strouhal number of dominant pressure fluctuations in the downstream of elbow is estimated to be around 0.45. In addition, the validity of the analytical tool was confirmed by comparing between analysis and experiment.

JAEA Reports

Study on assembly techniques and procedures for ITER tokamak device

Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi*; Ue, Koichi*; Shimizu, Katsusuke*; Onozuka, Masanori*

JAEA-Technology 2006-034, 85 Pages, 2006/06

JAEA-Technology-2006-034.pdf:9.18MB

The International Thermonuclear Experimental Reactor (ITER) tokamak is composed of many kinds of components. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of these components are required to be a high accuracy of $$pm$$3mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements. Based on the above background, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology concept based on the available knowledge and experiences of the installation of the large components, in order to improve the IT (International Team) design toward the more realistic one. As a result, we show the realistic tokamak assembly procedures and techniques to be able to assemble the large and heavy ITER tokamak with high accuracy. (1)Assembly and alignment of the toroidal field coil with high accuracy. (2)Simplification of the assembly procedures, and the jigs/tools and procedures to reduce the misalignment. (3)Assembly procedures and techniques for the vacuum vessel to reduce the weld distortion. (4)Supporting procedures and techniques of the vacuum vessel sector to prevent the toridal field coil from the deformation due to the dead weight of the vacuum vessel sector. (5)Datum points and lines for the required positions and assembly tolerances during tokamak assembly.

JAEA Reports

Applicability assessment of plug weld to ITER vacuum vessel by crack propagation analysis

Omori, Junji; Nakahira, Masataka; Takeda, Nobukazu; Shibanuma, Kiyoshi; Sago, Hiromi*; Onozuka, Masanori*

JAEA-Technology 2006-017, 134 Pages, 2006/03

JAEA-Technology-2006-017.pdf:15.96MB

In order to improve the fabricability of the vacuum vessel (VV) of International Thermonuclear Experimental Reactor (ITER), applicability of plug weld between VV outer shell and stiffening ribs/blanket support housings has been assessed using crack propagation analysis for the plug weld. The ITER VV is a double-wall structure of inner and outer shells with ribs and housings between the shells. For the fabrication of VV, ribs and housings are welded to outer shell after welding to inner shell. A lot of weld grooves should be adjusted for the outer shell weld. The plug weld can allow larger tolerance of weld groove gaps than ordinary butt weld. However, un-welded lengths parallel to outer shell surface remain in the plug weld region. It is necessary to evaluate the allowable un-welded length to apply the plug weld to ITER vacuum vessel fabrication. For the assessment the allowable un-welded lengths have been calculated by crack propagation analyses for the load conditions, conservatively assuming the un-welded region is a crack. The analyses have been carried out for typical inboard straight region and inboard upper curved region with maximum housing stress. The allowable cracks of ribs are estimated to be 8.8mm and 38mm for the rib and the housing, respectively, considering inspection error of 4.4mm. Plug welding for welding between outer shell and ribs/housings could be applicable.

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