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Journal Articles

Load and resistance factor design approach for seismic buckling of fast reactor vessels

Takaya, Shigeru; Sasaki, Naoto*; Asayama, Tai; Kamishima, Yoshio*

Mechanical Engineering Journal (Internet), 4(3), p.16-00558_1 - 16-00558_12, 2017/06

In this study, we developed a new design rule for the prevention of seismic buckling of vessels using the load and resistance factor design method to enable more rational vessel designs. The effectiveness of the new design rule was illustrated in comparison with the current provision.

Journal Articles

Development of core and structural materials for fast reactors

Asayama, Tai; Otsuka, Satoshi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 15 Pages, 2017/06

This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design.

Journal Articles

Development of simple estimation method for the influence of parameter uncertainty of probability distributions against evaluation result of probabilistic fracture mechanics

Okajima, Satoshi; Takaya, Shigeru; Asayama, Tai

Nippon Kikai Gakkai Rombunshu (Internet), 83(845), p.16-00434_1 - 16-00434_13, 2017/01

no abstracts in English

Journal Articles

Determination of in-service inspection requirements for fast reactor components using System Based Code concept

Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai

Nuclear Engineering and Design, 305, p.270 - 276, 2016/08

AA2016-0006.pdf:0.51MB

 Times Cited Count:2 Percentile:70.56(Nuclear Science & Technology)

In our previous study, we proposed a new process for determining the in-service inspection (ISI) requirements using the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, the ISI requirements for a reactor guard vessel (RGV) and core support structure (CSS) of a prototype sodium-cooled fast breeder reactor in Japan (Monju) were investigated using the proposed process. It was shown that both components had sufficient reliability even assuming unrealistic severe conditions. The failure occurrences of these components were practically eliminated. Hence, it was concluded that no ISI requirements were needed for these components. The proposed process is expected to contribute to the realization of effective and rational ISI by properly taking into account plant-specific features.

Journal Articles

Study on minimum wall thickness requirement for seismic buckling of reactor vessel based on system based code concept

Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai

Journal of Pressure Vessel Technology, 137(5), p.051802_1 - 051802_7, 2015/10

 Times Cited Count:2 Percentile:83.04(Engineering, Mechanical)

The minimum wall thickness required to prevent seismic buckling of a reactor vessel in a fast reactor is derived using the System Based Code (SBC) concept. One of the key features of SBC concept is margin optimization; to implement this concept, the reliability design method is employed, and the target reliability for seismic buckling of the reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation, such as distribution type, mean value, and standard deviation of random variables, are also prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Minimum wall thickness required to achieve the target reliability is evaluated, and is found to be less than that determined from a conventional deterministic design method. Furthermore, the influence of each random variable on the evaluation is investigated, and it is found that the seismic load has a significant impact.

Journal Articles

Development of creep-fatigue evaluation method for 316FR stainless steel

Nagae, Yuji; Takaya, Shigeru; Asayama, Tai

Journal of Pressure Vessel Technology, 137(4), p.041407_1 - 041407_5, 2015/08

 Times Cited Count:0 Percentile:100(Engineering, Mechanical)

Journal Articles

Determination of ISI requirements on the basis of system based code concept

Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai

Transactions of 23rd International Conference on Structural Mechanics in Reactor Technology (SMiRT-23) (USB Flash Drive), 10 Pages, 2015/08

In our previous study, a new process for determination of in-service inspection (ISI) requirements was proposed on the basis of the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, ISI requirements for a reactor guard vessel and a core support structure of the prototype sodium-cooled fast breeder reactor in Japan, Monju, were investigated according to the proposed process. The proposed process is expected to contribute to realize effective and rational ISI by properly taking into account plant-specific features.

Journal Articles

Application of the system based code concept to the determination of in-service inspection requirements

Takaya, Shigeru; Asayama, Tai; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Nakai, Satoru; Morishita, Masaki

Journal of Nuclear Engineering and Radiation Science, 1(1), p.011004_1 - 011004_9, 2015/01

A new process for determination of inservice inspection (ISI) requirements was proposed based on the System Based Code concept to realize effective and rational ISI by properly taking into account plant specific features. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other one on detectability of defects before they would grow to an unacceptable size in light of plant safety. If defect detection was not feasible, structural integrity evaluation would be required under sufficiently conservative hypothesis. The applicability of the proposed process was illustrated through an application to the existing prototype fast breeder reactor, Monju.

Journal Articles

Development of structural codes for JSFR based on the system based code concept

Asayama, Tai; Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Nagae, Yuji; Takaya, Shigeru; Onizawa, Takashi; Tsukimori, Kazuyuki; Morishita, Masaki

Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 6 Pages, 2014/07

This paper overviews the ongoing research and development as well as activities for codification of structural codes for the Japan Sodium Cooled Fast Reactor (JSFR). Not only the design and construction code which has been published and updated on a regular basis, codes on welding, fitness-for-service, leak-before-break evaluation as well as the guidelines for structural reliability evaluation are being developed. The basic strategy for the development is to fully take advantage of the favorable technical characteristics associated with sodium-cooled fast reactors; the codes will be developed based on the System Based Code concept. The above mentioned set of codes are planned to be published from the Japan Society of Mechanical Engineers in 2016.

Journal Articles

Elaboration of the system based code concept; Activities in JSME and ASME, 1; Overview

Asayama, Tai; Miyagawa, Takayuki*; Dozaki, Koji*; Kamishima, Yoshio*; Hayashi, Masaaki*; Machida, Hideo*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07

This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept. A brief introduction to the SBC concept is followed by the technical features of structural evaluation methodologies that are based on the SBC concept. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.

Journal Articles

Elaboration of the system based code concept; Activities in JSME and ASME, 4; Joint efforts of JSME and ASME

Asayama, Tai; Takaya, Shigeru; Morishita, Masaki; Schaaf, F.*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 9 Pages, 2014/07

This paper describes the ongoing activities at the Joint Task Group for System Based Code established in 2012 by the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) in the ASME Boiler and Pressure Vessel Code Committee. The Joint Task Group aims at developing alternative rules for ASME Boiler and Pressure Vessel Code Section XI Division 3, inservice inspection requirements for liquid metal reactors. The alternative rules will be developed based on the System Based Code concept which was originally proposed in Japan and is being elaborated both in JSME and ASME.

Journal Articles

Elaboration of the system based code concept; Activities in JSME and ASME, 2; Development of evaluation tools based on LRFD

Machida, Hideo*; Asayama, Tai; Watanabe, Taigo*; Hojo, Kiminobu*; Hayashi, Masaaki*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07

In the System Based Code (SBC), a reliability target is defined according to the importance of components (risk and/or failure probability), and grade of material, design, manufacture and maintenance are chosen to satisfy the reliability target. Therefore, reliability evaluation of components plays the important role of the concept of SBC. Until now, the LRFD methods were developed for burst due to internal pressure, plastic collapse due to membrane and bending stress, fatigue, limit load assessment of flawed pipe, and buckling of thin wall cylinder. This paper describes the action plans of development of the reliability assessment methods and an examination results up to date.

Journal Articles

Development of creep-fatigue evaluation method for modified 9Cr-1Mo steel

Takaya, Shigeru; Nagae, Yuji; Asayama, Tai

Journal of Pressure Vessel Technology, 136(3), p.031404_1 - 031404_8, 2014/06

 Times Cited Count:4 Percentile:67.17(Engineering, Mechanical)

This paper describes a creep-fatigue evaluation method for modified 9Cr-1Mo steel, which has been newly included in the 2012 edition of the JSME code for design and construction of fast reactors. In this method, creep and fatigue damages are evaluated on the basis of Miner's rule and the time fraction rule, respectively, and the linear summation rule is employed as the failure criterion. The conservativeness of this method without design factors was investigated using material test results, and it was shown that the time fraction approach can conservatively predict failure life.Comparison with the modified ductility exhaustion method, which is known to have good failure life predictability in material test results, shows that the time fraction approach predicts failure lives to be shorter in long-term strain hold conditions. These results confirm that the creep-fatigue evaluation method in the JSME FRs code has implicit conservatism.

Journal Articles

Evaluation of reduction in creep strength based on fracture energy in CSEF steels

Nagae, Yuji; Asayama, Tai

Advances in Materials Technology for Fossil Power Plants; Proceedings from the 7th International Conference (EPRI 2013), p.1304 - 1312, 2014/01

Journal Articles

Development of 2012 edition of JSME code for design and construction of fast reactors, 5; Creep-fatigue evaluation method for 316FR stainless steel

Nagae, Yuji; Takaya, Shigeru; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 6 Pages, 2013/07

Journal Articles

Extrapolation of creep strength by fracture energy for 316FR stainless steel at 823 K

Nagae, Yuji; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 7 Pages, 2013/07

Journal Articles

Development of 2012 edition of JSME code for design and construction of fast reactors, 6; Design margin assessment for the new materials to the rules

Ando, Masanori; Watanabe, Sota*; Kikuchi, Koichi*; Otani, Tomomi*; Sato, Kenichiro*; Tsukimori, Kazuyuki; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 11 Pages, 2013/07

New 2012 edition of JSME code for design and construction of fast reactors (FRs code) was published by Japan society of mechanical engineers (JSME). Main topic of the current JSME FRs code 2012 edition is registration of the two new materials, 316FR and Mod.9Cr-1Mo steel. The design margins for the new materials to the rules for the components and piping serviced at elevated temperature described in the JSME FRs code were assessed. To confirm the design margins, a series of the assessment program for the new materials to the conventional design rules was performed using the evaluation of the experimental data and finite element analysis. Through these assessments, the enough design margins for new materials to the rules were confirmed.

Journal Articles

Development of 2012 edition of JSME code for design and construction of fast reactors, 4; Creep-fatigue evaluation method for modified 9CR-1MO steel

Takaya, Shigeru; Nagae, Yuji; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 10 Pages, 2013/07

This paper describes a creep-fatigue evaluation method for modified 9Cr-1Mo steel, which has been newly included in the 2012 edition of the JSME code for design and construction of fast reactors. In this method, fatigue and creep damages are evaluated on the basis of Miner's rule and the time fraction rule, respectively, and the linear summation rule is employed as the failure criterion. Investigations using material test results are conducted, which show that the time fraction approach can conservatively predict failure life if margins on the initial stress of relaxation and the stress relaxation rate are embedded. In addition, the conservatism of prediction tends to increase with time to failure. Comparison with the modified ductility exhaustion method, which is known to have good failure life predictability in material test results, shows that the time fraction approach predicts failure lives to be shorter in longterm strain hold conditions.

Journal Articles

Study on minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling by system based code

Takaya, Shigeru; Watanabe, Daigo*; Yokoi, Shinobu*; Kamishima, Yoshio*; Kurisaka, Kenichi; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 6 Pages, 2013/07

In this paper, minimum wall thickness requirement of reactor vessel of fast reactor for seismic buckling is discussed on the basis of the System Based Code (SBC) concept. One of key concepts of SBC is the margin optimization. To implement this concept, reliability design method is employed, and the target reliability for seismic buckling of reactor vessel is derived from nuclear plant safety goals. Input data for reliability evaluation such as distribution type, mean value and standard deviation of random variable are prepared. Seismic hazard is considered to evaluate uncertainty of seismic load. Wall thickness needed to achieve the target reliability is evaluated, and as a result, it is shown that the minimum wall thickness can be reduced from that required by a deterministic design method.

Journal Articles

Development of 2012 edition of JSME code for design and construction of fast reactors, 3; Development of the material strength standard of modified 9Cr-1Mo steel

Onizawa, Takashi; Nagae, Yuji; Takaya, Shigeru; Asayama, Tai

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07

109 (Records 1-20 displayed on this page)