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Journal Articles

Study on a unified criterion for preventing plastic strain accumulation due to long distance travel of temperature distribution

Okajima, Satoshi; Wakai, Takashi; Kawasaki, Nobuchika

Mechanical Engineering Journal (Internet), 4(5), p.16-00641_1 - 16-00641_11, 2017/10

Journal Articles

Study on unified criteria for preventing plastic strain accumulation due to thermal stress with long range travel

Okajima, Satoshi; Wakai, Takashi; Kawasaki, Nobuchika

Proceedings of International Conference on Asia-Pacific Conference on Fracture and Strength 2016 (APCFS 2016) (USB Flash Drive), p.269 - 270, 2016/09

Journal Articles

Development of seismic isolation systems for sodium-cooled fast reactors in Japan

Kawasaki, Nobuchika; Watakabe, Tomoyoshi; Wakai, Takashi; Yamamoto, Tomohiko; Fukasawa, Tsuyoshi*; Okamura, Shigeki*

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 8 Pages, 2016/07

Sodium-cooled Fast Reactors (SFRs) have components with thinner walls as compared with light water reactors, although Japan is an earthquake-prone country. Thus, seismic isolation systems have been conventionally employed in SFR system design to reduce seismic forces on the systems in Japan. Implementation of seismic design in the reactor core and buckling design in the reactor vessel requires 8 Hz (or less) vertical frequency's isolation system being applied. This paper introduces three isolation concepts to achieve the frequency. The isolation systems, which enable vertical 8 Hz natural frequency, comprise thicker laminated rubber bearings (TRBs). By combining coned disk springs with TRBs, vertical natural frequency is in a range from roughly 3 Hz to 5 Hz. Combining pneumatic springs to RBs and adding the rocking suppression system, vertical natural frequency becomes under 1 Hz. All isolation systems need horizontal damping like oil dampers. A vertical 8 Hz isolation system with TRBs and oil dampers is under development in Japan as a principal isolation concept. The reasons of choosing this system are its simplicity and the number of developing issues. Since TRBs and oil dampers are basic isolation elements, they can be applied to other isolation systems. The response acceleration of 5 Hz vertical isolation is 50% of that of 8 Hz based on the analytical survey. A series of static tests of coned disk springs was carried out to confirm design equations. Based on these knowledge, 5 Hz vertical isolation system with TRBs and the coned disk springs can be designed. The response acceleration of 1 Hz vertical isolation is 10% of that of 8 Hz. A rocking suppression system was studied in the past, and the further simplification of this system is the largest challenge for this concept. These three isolation concepts are isolation candidates for SFRs in Japan. To obtain enough seismic margins for each plant site, these isolation systems need to be developed.

Journal Articles

Development on rubber bearings for sodium-cooled fast reactor, 3; Ultimate properties of a half scale thick rubber bearings based on breaking test

Fukasawa, Tsuyoshi*; Okamura, Shigeki*; Yamamoto, Tomohiko; Kawasaki, Nobuchika; Hirotani, Tsutomu*; Moriizumi, Eriko*; Sakurai, Yu*; Masaki, Nobuo*

Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 10 Pages, 2016/07

Half-scale thick rubber bearing to investigate ultimate properties application for a Sodium-cooled-Fast-Reactor. The fundamental restoring-force characteristics of the thick rubber bearings has been already cleared through the static loading tests using a half-scale thick rubber bearing, 800 mm in diameter. However, variations of the restoring force characteristics and ultimate properties have not been obtained yet. The purpose of this paper is to indicate the variation of the stiffness and damping ratio concerning restoring force characteristics and the breaking strain or stress as ultimate properties through static loading tests using the half-scale thick rubber bearings.

Journal Articles

Parametric design study about seismic isolation system for fast reactor JSFR

Kawasaki, Nobuchika; Yamamoto, Tomohiko; Fukasawa, Tsuyoshi*; Okamura, Shigeki*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 9 Pages, 2015/07

Japanese seismic conditions are getting severer and natural frequencies of components are getting lower due to the enlargements of components' size, therefore response accelerations and buckling margins of reactor vessels were parametrically surveyed with attention to thicknesses, diameters, and isolation frequencies for reviewing necessary isolation specification. RV installed floor responses and buckling margins were calculated based on this seismic condition. Expansion characteristic of isolation system was evaluated by parametric acceleration response analyses. Japanese seismic design condition may become severer than present one, and a natural frequency of main component may decrease. However based on the buckling margin with present plant specifications and the expansion characteristic of isolation system, the advanced isolation system with 8Hz vertical natural frequency was selected as the isolation system of JSFR at still present occasion.

Journal Articles

Development on rubber bearings for sodium-cooled fast reactor, 1; Examination plan

Yamamoto, Tomohiko; Kawasaki, Nobuchika; Fukasawa, Tsuyoshi*; Okamura, Shigeki*; Somaki, Takahiro*; Samejima, Yusuke*; Masaki, Nobuo*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 7 Pages, 2015/07

Since a SFR (Sodium-cooled Fast Reactor) has thin-walled component structures, a seismic isolation system is employed to mitigate the seismic force. Seismic isolation system for applying the SFR consists of the laminated rubber bearing considering characteristics of SFR structures. This paper describes a basic mechanical characteristic examination with a 1/8 scale model and a characterization examination plan of half-scale laminated rubber.

Journal Articles

Development on rubber bearings for sodium-cooled fast reactor, 2; Fundamental characteristics of half-scale rubber bearings based on static test

Fukasawa, Tsuyoshi*; Okamura, Shigeki*; Yamamoto, Tomohiko; Kawasaki, Nobuchika; Somaki, Takahiro*; Sakurai, Yu*; Masaki, Nobuo*

Proceedings of 2015 ASME Pressure Vessels and Piping Conference (PVP 2015) (Internet), 10 Pages, 2015/07

This paper described the results of static loading tests using a half-scale rubber bearing model to investigate the fundamental characteristics such as restoring force of a rubber bearing applied to a Sodium-Cooled-Fast-Reactor (SFR). Since the SFR has thin-walled structures, a seismic isolation system is employed to mitigate the seismic force. The static loading tests were performed using the half-scale rubber bearing with a diameter of 800 mm in the range which exceeds a linear limit of horizontal direction and a yield stress of vertical direction to investigate the horizontal and vertical of each stiffness and damping ratio. The fundamental characteristic of rubber bearing employed to the SFR and the validity of a design formula became clear through the static tests.

Journal Articles

Design study for reactor system of fast reactor JSFR; Concept of reactor system

Kawasaki, Nobuchika; Sakamoto, Yoshihiko; Eto, Masao*; Taniguchi, Yoshihiro*; Kamishima, Yoshio*

Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.760 - 769, 2015/05

The Japan Sodium-cooled Fast Reactor, JSFR, is currently under conceptual study. The concept of JSFR's reactor system is a compact reactor system to avoid excessive increase of reactor vessel diameter with structural and fluid integrities. To realize this concept, single rotating plug with advanced refueling system is adopted. Advanced refueling system consists of column type Upper Internal Structure and pantograph type Fuel Handling Machine. To realize structural and fluid integrities, top entry piping, sodium dam and flow block/guide structures are adopted. Structural integrities against seismic displacement or thermal stress and fluid integrities against vortex cavitations or cover gas entrainment can be ensured with these designs.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 3; Progress of component design

Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi*; Eto, Masao*; Miyagawa, Takayuki*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In the frame work of generation IV international forum, safety design criteria and safety design guideline for the generation IV sodium-cooled fast reactors have been developing. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC. In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX was modified for the primary heat exchanger, which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator, protective wall tube type design is under investigation as an option with less R&D risks.

Journal Articles

Comparison and assessment of the creep-fatigue evaluation methods with notched specimens made of Mod.9Cr-1Mo steel

Ando, Masanori; Hirose, Yuichi*; Karato, Takanori*; Watanabe, Sota*; Inoue, Osamu*; Kawasaki, Nobuchika; Enuma, Yasuhiro*

Journal of Pressure Vessel Technology, 136(4), p.041406_1 - 041406_10, 2014/08

 Times Cited Count:7 Percentile:53(Engineering, Mechanical)

To compare and assess the creep-fatigue life evaluation methods for stress concentration point, a series of creep-fatigue test was performed with notched specimens made of Mod.9Cr-1Mo steel. Mechanical creep-fatigue tests and thermal creep-fatigue test were performed. A series of Finite Element Analysis was also carried out to predict the number of cycles to failure by the several creep-fatigue life evaluation methods. Then these predictions were compared with the test results. Several types of evaluation methods such are stress redistribution locus (SRL) method, simple elastic follow-up method and the methods described in the JSME FRs code were applied. Through the comparisons, it was appeared that SRL method gave rational conservative prediction of the creep-fatigue life for all conditions tested in this study. The JSME FRs code gave an evaluation over 70 times conservative lives comparing with the test results.

Journal Articles

Development of constitutive models for fast reactor design

Tsukimori, Kazuyuki; Iwata, Koji*; Kawasaki, Nobuchika*; Okajima, Satoshi; Yada, Hiroki; Kasahara, Naoto*

Nuclear Engineering and Design, 269, p.23 - 32, 2014/04

 Times Cited Count:1 Percentile:88.57(Nuclear Science & Technology)

R&D to enable a practical fast breeder reactor plant is proceeding in Japan, which is called "FaCT" (Fast reactor Cycle Technology development). One of the key issues of R&D is to realize a reasonably compact reactor vessel by eliminating the wall protection equipment which is installed inside the vessel in order to reduce thermal loading in the conventional design. Most important concern is the amount of the inelastic strain of the vessel accumulated around the liquid sodium surface which moves upward and downward cyclically with start-up and shut-down. The aim of this study is to develop rational constitutive models that enable prediction of this kind of complex inelastic behaviors precisely and to prepare the design guide based on inelastic analysis. We developed a high accuracy plasticity model and a simplified plasticity model and valuated them by organized experiments.

Journal Articles

Stress mitigation design of a tubesheet by considering the thermal stress inducement mechanism

Ando, Masanori; Takasho, Hideki*; Kawasaki, Nobuchika; Kasahara, Naoto*

Journal of Pressure Vessel Technology, 135(6), p.061207_1 - 061207_10, 2013/12

 Times Cited Count:5 Percentile:66.27(Engineering, Mechanical)

The stress generation mechanism of a tubesheet was revealed through finite element analysis. Semi-spherical tubesheet models were investigated for the first survey of the thermal stress mechanism. The calculated results of the semi-spherical tubesheet model indicated an extensive peak stress around the outermost hole. The recognized thermal stress mechanism of a semi-spherical tubesheet is summarized, and on the basis of the stress generation mechanism, we proposed a stress-mitigated tubesheet, a center-flattened spherical tubesheet (CFST), as an improved configuration. The stress generation mechanism of the CFST was also desicibed.

Journal Articles

Verification of the estimation methods of strain range in notched specimens made of Mod.9Cr-1Mo steel

Ando, Masanori; Hirose, Yuichi*; Date, Shingo*; Watanabe, Sota*; Enuma, Yasuhiro*; Kawasaki, Nobuchika

Journal of Pressure Vessel Technology, 134(6), p.061403_1 - 061403_12, 2012/12

 Times Cited Count:3 Percentile:76.12(Engineering, Mechanical)

To verify the methods of estimating strain range for discontinues structures, low cycle fatigue tests were carried out with notched specimens. All the specimens were made of Mod.9Cr-1Mo. Displacement control fatigue tests and thermal fatigue tests were performed by ordinary uni-axial push pull test machine and equipment generating the thermal gradient in the notched plate by induction heating. Elastic and inelastic finite element analysis were also performed to estimate strain range for predicting fatigue life. Then these predictions were compared with the test results. Several methods such as stress reduction locus (SRL) method, simple elastic follow-up (SEF) method, Neuber's law and the procedures employed by elevated temperature design codes were applied. Through these comparisons, the applicability and conservativeness of these strain range estimation methods are discussed.

Journal Articles

Experimental investigation of strain concentration evaluation based on the stress redistribution locus method

Isobe, Nobuhiro*; Kawasaki, Nobuchika; Ando, Masanori; Sukekawa, Masayuki*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 11 Pages, 2012/00

Evaluation of local strain at structural discontinuities is an important technology in high temperature design of fast reactors because the failure mode in high temperature fatigue or creep fatigue damage is usually crack initiation and growth from such a locally high strained area. A rationalized strain concentration evaluation method was discussed experimentally in this study. The stress redistribution locus (SRL) method had been proposed to improve the accuracy of local stress and strain evaluation for structural discontinuities. High temperature fatigue tests of circumferentially notched specimens were conducted accompanying with local strain measurement by a capacitance type strain gage. Measured strain was compared with the prediction by the SRL method and the applicability of the method is discussed.

Journal Articles

Limitation study for cyclic hardening recovery for 316FR stainless steel derived from long-term holding with elevated temperature

Okajima, Satoshi; Kawasaki, Nobuchika*; Fukahori, Takuya*; Kikuchi, Koichi*; Kasahara, Naoto

Dai-49-Kai Koon Kyodo Shimpojiumu Koen Rombunshu, p.85 - 89, 2011/11

no abstracts in English

Journal Articles

Development of constitutive models for fast reactor design

Tsukimori, Kazuyuki; Iwata, Koji*; Kawasaki, Nobuchika*; Kasahara, Naoto*

Transactions of 21st International Conference on Structural Mechanics in Reactor Technology (SMiRT-21) (CD-ROM), 8 Pages, 2011/11

Journal Articles

Conceptual design study for the demonstration reactor of JSFR, 4; Structural design of reactor vessel

Kawasaki, Nobuchika; Okamura, Shigeki*; Sawa, Naoki*; Sakamoto, Yoshihiko; Negishi, Kazuo

Proceedings of 19th International Conference on Nuclear Engineering (ICONE-19) (CD-ROM), 7 Pages, 2011/10

Japan Sodium-Cooled Fast Reactor adopts an compacted hot reactor vessel concept. From the point of structural designs to ensure both seismic design and elevated temperature design is important. In this study, based on a common conservative seismic loading condition considered with the Niigata-ken Chuetsu-oki Earthquake, seismic evaluations were carried out, the thicknesses of reactor vessels of 750 MWe and 500 MWe output plants were determined. For both plants 50 mm was selected as the thickness, and ensured buckling evaluation margins were more than 2.4. From the point of seismic design, the difference of plant output was negligible. With the condition of 50 mm thickness of reactor vessel, thermal integrities were evaluated. For three plant start-up conditions which were 2.2, 3.2, and 4.3 days, thermal ratcheting and creep-fatigue damage were evaluated. As a result plant start-up period needed more than 3.2 days for both 750 MWe and 500 MWe output plants. Caused thermal stress were the nearly same for both plants, therefore from the point of thermal design, the difference of plant output was negligible.

Journal Articles

Comparison of creep-fatigue evaluation methods with notched specimens made of Mod.9Cr-1Mo steel

Ando, Masanori; Hirose, Yuichi*; Karato, Takanori*; Watanabe, Sota*; Inoue, Osamu*; Kawasaki, Nobuchika; Enuma, Yasuhiro*

Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 11 Pages, 2011/07

Journal Articles

An Evaluation method for plastic buckling of cantilever type pipes controlled by displacement loads, 1; Proposal of the estimation method and the criterion

Ando, Masanori; Tezuka, Taiji*; Nakamura, Toshio*; Okawa, Tomohiro*; Enuma, Yasuhiro*; Kawasaki, Nobuchika; Tsukimori, Kazuyuki

Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 9 Pages, 2011/07

Journal Articles

An Evaluation method for plastic buckling of cantilever type pipes controlled by displacement loads, 2; Verification of proposal method by buckling test

Ando, Masanori; Tezuka, Taiji*; Nakamura, Toshio*; Okawa, Tomohiro*; Enuma, Yasuhiro*; Kawasaki, Nobuchika; Tsukimori, Kazuyuki

Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 8 Pages, 2011/07

83 (Records 1-20 displayed on this page)