Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 51

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Development of the buckling evaluation method for large scale vessels in fast reactors by the testing of Grade 91 steel and austenitic stainless steel vessels subjected to horizontal and cyclic vertical loading

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Okajima, Satoshi

Proceedings of ASME 2023 Pressure Vessels and Piping Conference (PVP 2023) (Internet), 8 Pages, 2023/07

Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Gr.91 and austenitic stainless steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively.

Journal Articles

Development of the buckling evaluation method for large scale vessel in fast reactors by the testing of austenitic stainless steel vessel with severe initial imperfection subjected to horizontal and vertical loading

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Okajima, Satoshi

Proceedings of ASME 2022 Pressure Vessels and Piping Conference (PVP 2022) (Internet), 9 Pages, 2022/07

Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of austenitic stainless steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. Moreover, a series of finite element analyzes to confirm the applicability of the evaluation method in 2.25Cr-1Mo steel and up to 650 $$^{circ}$$C were conducted.

Journal Articles

Development of the buckling evaluation method for large scale vessel by the testing of Gr.91 vessel subjected to horizontal and cyclic vertical loading

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07

Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of cyclic axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. Moreover, a series of finite element analyzes using a model with residual stress due to welding revealed that the effect of residual stress on buckling strength is negligible in the evaluation method.

Journal Articles

Study on the predictive evaluation method for loads acting on roof and sidewall of cylindrical tank in nonlinear sloshing based on simplified equations

Ikesue, Shunichi*; Morita, Hideyuki*; Ishii, Hidekazu*; Sago, Hiromi*; Yokoi, Shinobu*; Yamamoto, Tomohiko

Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 8 Pages, 2021/07

In this paper, a new method is proposed for the nonlinear sloshing condition of a cylindrical tank, which can evaluate the vertical load acting on the roof and the horizontal load acting on the sidewall. This method is a combination of simplified equations for the liquid surface level and velocity proposed in the past study and the new pressure model modified from the existing model. A long calculation time as CFD analysis is not needed, because this method is consisted of simplified equations. The validity of this method was confirmed by comparing them with the CFD and the test. And future issues on the improvement of this method were clarified from the result.

Journal Articles

Development of the buckling evaluation method for large scale vessel by the testing of Gr.91 vessel subjected to vertical and horizontal loading

Okafuji, Takashi*; Miura, Kazuhiro*; Sago, Hiromi*; Murakami, Hisatomo*; Ando, Masanori; Miyazaki, Masashi

Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 9 Pages, 2020/08

Buckling evaluation methods capable of evaluating elasto-plastic buckling under axial compression, bending, and shear loads are required for cylindrical vessels of fast reactors to cope with thinning due to increasing diameter and application to the seismic isolation design against huge seismic ground motion. In this study, in order to confirm the applicability of the proposal evaluation method, several buckling tests and FE analyses were carried out using the specimens made of Modified 9Cr-1Mo steel. The buckling modes and strength data in the load region where the interaction of axial compression, bending and shear buckling could occur were examined. As a result, it was confirmed that the proposal evaluation method estimated the buckling load in the tests conservatively. In addition, buckling strength evaluated by elasto-plastic buckling analysis had good accuracy compared to each test result by considering the stress-strain relationship and imperfection of test specimen.

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

Nihon 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

Nihon 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*

Nihon 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, 2011/10

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, 2011/10

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, 2011/10

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:70.51(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.

51 (Records 1-20 displayed on this page)