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Journal Articles

Study on fracture behaviour of through-wall cracked elbow under displacement control load

Machida, Hideo*; Koizumi, Yu*; Wakai, Takashi; Takahashi, Koji*

Nippon Kikai Gakkai M&M 2019 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), p.OS1307_1 - OS1307_5, 2019/11

This paper describes the fracture test and fracture analysis of a pipe under displacement control load. In order to grasp the fracture behavior of the circumferential through-wall cracked pipe, which is important in evaluating the feasibility of leak before break (LBB) in sodium cooled reactor piping, a fracture test in case of a circumferential throughwall crack in the weld line between an elbow and a straight pipe was carried out. From this test, it was found that no pipe fracture occurs in the displacement control loading condition even if a large circumferential through-wall crack (180$$^{circ}$$) was assumed. The fracture analysis of the pipe was carried out using Gurson's parameters set based on the tensile test results of the tested pipe material. The analytic results agree well with the test results, and it was found that it will be possible to predict the fracture behavior of sodium cooled reactor piping.

Journal Articles

Effect of local plastic component on crack opening displacement and on J-integral of a circumferential penetrated crack

Machida, Hideo*; Arakawa, Manabu*; Wakai, Takashi

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

This paper describes the effect of local plastic component on J-integral and crack opening displacement (COD) evaluation of a circumferential penetrated crack, applicable to the leak before break (LBB) assessment for sodium cooled fast reactor (SFR) components. J-integral COD evaluation methods are generally formulated as a summation of elastic and plastic components, and so far many evaluation formulae based on these two components have been proposed. However, strictly, the plastic component consists of local plastic and fully plastic components. Many of the conventional evaluation methods often consider only the fully plastic component as the plastic component. The reason for this is that the effect of the local plastic component is much smaller than that of the fully plastic component excluding materials with extremely small work hardening. In contrast, for materials with high yield stress and small work hardening, such as modified 9Cr-1Mo steel which is one of the candidate materials for SFR piping, the effect of the local plastic component on J-integral and COD cannot be ignored. Therefore, the authors propose formulae taking the effect of local plastic component on J-integral and COD into account, based on finite element analysis (FEA) results, so that it is easy to apply to crack evaluation. The formulae will be employed in the guidelines on LBB assessment for SFR components published from Japan Society of Mechanical Engineers (JSME).

Journal Articles

Improvement of penetrate crack length evaluation method for LBB assessment of sodium-cooled fast reactor components

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*

Nippon Kikai Gakkai 2018-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2018/09

According to the fitness for service code of Sodium-Cooled fast Reactor (SFR), the volumetric tests as in-service inspection can be replaced with continuous leak monitoring, where the Leak Before Break (LBB) is demonstrated, because the primary stress caused by internal pressure is not significant in SFR components. Basically, if the detectable crack length and the penetrated crack length are sufficiently smaller than the unstable critical crack length, it can be concluded that LBB is successfully demonstrated. The authors had already proposed a simplified method to calculate the penetrated crack length both of the circumferential and axial cracks in the pipe as a function of pipe geometry, fatigue crack growth characteristics and loading conditions. However, some problems in the method have been pointed out in the process of the reviewing by the JSME code committee. This study describes an improved method to calculate the penetrated crack length.

Journal Articles

Development of a crack opening displacement assessment procedure considering change of compliance at a crack part in thin wall pipes made of modified 9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Yanagihara, Seiji*; Suzuki, Ryosuke*; Matsubara, Masaaki*

Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07

This paper studies crack opening displacement (COD) evaluation methods used in Leak-Before-Break (LBB) assessment of Sodium cooled Fast Reactor (SFR) pipe. For SFR pipe, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. The sodium pipes are made of ASME Gr.91 (modified 9Cr-1Mo steel). Thickness of the pipes is small, because the internal pressure is very small. Modified 9Cr-1Mo steel has a relatively large yield stress and small work hardening coefficient comparing to the austenitic stainless steels which are currently used in the conventional plants. In order to assess the LBB behavior of the sodium pipes made of modified 9Cr-1Mo steel, the coolant leak rate from a through wall crack must be estimated properly. Since the leak rate is strongly related to the crack opening displacement (COD), an appropriate COD assessment method must be established to perform LBB assessment. However, COD assessment method applicable for SFR pipes - having thin wall thickness and made of small work hardening material - has not been proposed yet. Thus, a COD assessment method applicable to such a pipe was proposed in this study. In this method, COD was calculated by classifying the components of COD; elastic, local plastic and fully plastic. In addition, the verification of this method was performed by comparing with the results of a series of four-point bending tests using modified 9Cr-1Mo steel pipe having a circumferential through wall notch. As a result, in some cases, COD were over-estimated especially for large cracks. Although the elastic component of COD is still over-estimated for large cracks, leak evaluation from small cracks is much more important in LBB assessment. Therefore, this study recommends that only the elastic component of COD should be adopted in LBB assessment of SFR pipes.

Journal Articles

Proposal of simplified J-integral evaluation method for a through wall crack in SFR pipe made of Mod.9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Kikuchi, Koichi*

Proceedings of ASME Symposium on Elevated Temperature Applications of Materials for Fossil, Nuclear, and Petrochemical Industries, 7 Pages, 2018/04

A simplified J-integral evaluation method applicable to unstable failure analysis in Leak Before Break (LBB) assessment of Sodium-cooled Fast Reactor (SFR) in Japan was proposed. Mod.9Cr-1Mo steel is supposed to be a candidate material for the coolant systems of SFR in Japan. This steel has relatively high yield strength and poor fracture toughness comparing to those of conventional austenitic stainless steels. In addition, SFR pipe has small thickness and large diameter. As a J-integral evaluation method for circumferential through-wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of SFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elasto-plastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to propose a J-integral evaluation method applicable to such loading conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed. The method enables to conduct 2 parameter failure analysis using J-integral without any fracture mechanics knowledge.

Journal Articles

Proposal on LBB evaluation conditions for sodium cooled fast reactor pipes and effects of pipe parameters

Yada, Hiroki; Takaya, Shigeru; Wakai, Takashi; Nakai, Satoru; Machida, Hideo*

Nippon Kikai Gakkai Rombunshu (Internet), 84(859), p.17-00389_1 - 17-00389_15, 2018/03

no abstracts in English

Journal Articles

Superconductivity in repulsively interacting fermions on a diamond chain; Flat-band-induced pairing

Kobayashi, Keita*; Okumura, Masahiko; Yamada, Susumu; Machida, Masahiko; Aoki, Hideo*

Physical Review B, 94(21), p.214501_1 - 214501_7, 2016/12

 Times Cited Count:32 Percentile:16.61(Materials Science, Multidisciplinary)

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

A Study on evaluation method of penetrate crack length for LBB assessment of fast reactor pipes

Wakai, Takashi; Machida, Hideo*; Sato, Kenichiro*

Nippon Kikai Gakkai M&M 2015 Zairyo Rikigaku Kanfarensu Koen Rombunshu (Internet), 3 Pages, 2015/11

This paper describes a through-wall crack length evaluation procedure applicable to Leak Before Break (LBB) assessment of Japan Sodium cooled Fast Reactor (JSFR) pipes made of Mod.9Cr-1Mo steel. In LBB assessment of JSFR pipes, it is required to calculate virtual through-wall crack length, though the crack growth is quite small under design condition. Generally, it is known that the through-wall crack length depends on loading condition, namely the load ratio between tensile and bending and that the length under pure bending load condition is largest. This study proposes a simplified method to evaluate the through-wall crack length both for axial and circumferential cracks as a function of load ratio and fatigue crack growth characteristics. Using the method, through-wall crack length can be predicted as far as we know the loading condition and material properties.

Journal Articles

Development of an unstable failure analysis procedure considering change of compliance at a crack part of SFR pipes

Wakai, Takashi; Machida, Hideo*; Yoshida, Shinji*; Yanagihara, Seiji*; Suzuki, Ryosuke*; Matsubara, Masaaki*; Enuma, Yasuhiro

Engineering Failure Analysis, 56, p.484 - 500, 2015/10

 Times Cited Count:1 Percentile:83.84(Engineering, Mechanical)

Journal Articles

J-integral evaluation method for a through wall crack in thin-walled large diameter pipes made of Mod.9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Sato, Kenichiro*

Nippon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09

This paper describes a J-integral evaluation procedure applicable to unstable failure analysis for a circumferential through wall crack in a pipe. Japan Sodium cooled Fast Reactor (JSFR) pipes are made of Mod.9Cr-1Mo steel. The fracture toughness of the material is inferior to that of conventional austenitic stainless steels. In addition, JSFR pipe has small thickness and large diameter and displacement controlled load is predominant. Therefore, the load balance in such piping system changes by crack extension and 2 parameter method using J-integral is applicable to unstable failure analysis for the pipes under such conditions. As a J-integral evaluation method for circumferential through wall crack in a cylinder, EPRI has proposed a fully plastic solution method. However, the geometry of JSFR pipe and material characteristics of Mod.9Cr-1Mo steel exceed the applicable range of EPRI's method. Therefore, a series of elastic, elastoplastic and plastic finite element analyses (FEA) were performed for a pipe with a circumferential through-wall crack to establish a J-integral evaluation method applicable to such conditions. J-integrals obtained from the FEA were resolved into elastic, local plastic and fully plastic components. Each component was expressed as a function of analytical parameter, such as pipe geometries, crack size, material characteristics and so on. As a result, a simplified J-integral evaluation method was proposed.

Journal Articles

Requirements for fracture toughness to satisfy LBB behavior of a pipe made of high chromium steel

Machida, Hideo*; Wakai, Takashi; Sato, Kenichiro*

Nippon Kikai Gakkai 2015-Nendo Nenji Taikai Koen Rombunshu (DVD-ROM), 5 Pages, 2015/09

The volumetric test for piping in a sodium cooled fast reactor (SFR) is difficult from the poor accessibility. Detection of a crack, therefore, is difficult before its penetration of a pipe wall, an SFR has a strategy to detect sodium leakage from a through wall crack before fracture of a pipe. Plant safety is ensured by shutting down a plant as soon as possible to detect small quantity of sodium leakage even if a crack penetrates a pipe wall. Consequently, it is important to ensure establishment of leakage-before-break (LBB) in this strategy. Effects of fracture resistance curve on fracture strength of a cracked pipe made of high chromium steel (Mod. 9Cr-1Mo steel), which is one of the candidates for fast reactor piping material, are evaluated in this study; and requirements for fracture resistance curve to achieve the LBB were proposed.

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

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, 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

Elaboration of the system based code concept; Activities in JSME and ASME, 3; Guidelines on structural reliability evaluation for FBR

Takaya, Shigeru; Machida, Hideo*; Kamishima, Yoshio*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 4 Pages, 2014/07

This paper describes the outline of the guidelines on structural reliability evaluation for the passive components of the fast breeder reactor (FBR). The guidelines are now being prepared by the task force for the system based code in the Japan society of mechanical engineers in order to contribute to reducing differences in evaluated structural reliability by evaluators. They consist of five chapters, which are "General rules", "Reliability evaluation", "Failure scenario setting", "Modeling", and "Failure probability calculation", respectively. Details of each chapter are explained.

Journal Articles

Demonstration of Leak-Before-Break in Japan sodium cooled fast reactor (JSFR) pipes

Wakai, Takashi; Machida, Hideo*; Yoshida, Shinji*; Xu, Y.*; Tsukimori, Kazuyuki

Nuclear Engineering and Design, 269, p.88 - 96, 2014/04

 Times Cited Count:12 Percentile:21.68(Nuclear Science & Technology)

Journal Articles

Development of limit state design for fast reactor by system based code

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

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

Limit state design was newly developed in order to apply the margin exchange which is one of the innovative concepts of the System Based Code (SBC). It was shown that limit state design method is applicable to plant design instead of current design criteria. In this report, working example of a reactor vessel of a Fast Reactor subject to thermal load is conducted to demonstrate this concept. As the result allowable stress was increased by changing the acceptance criteria from current design criteria to limit state design criteria.

Journal Articles

Development of leak-before-break assessment method for Japan Sodium cooled Fast Reactor pipe, 1; Crack opening displacement assessment of thin wall pipes made of modified 9Cr-1Mo steel

Wakai, Takashi; Machida, Hideo*; Arakawa, Manabu*; Yoshida, Shinji*; Enuma, Yasuhiro

Journal of Pressure Vessel Technology, 135(1), p.011401_1 - 011401_9, 2013/02

 Times Cited Count:1 Percentile:89.56(Engineering, Mechanical)

This was carried out to establish crack opening displacement (COD) evaluation methods used in Leak-Before-Break (LBB) assessment of sodium pipes of the Japan Sodium cooled Fast Reactor (JSFR). A COD assessment method applicable to thin walled large diameter pipe made of modified 9Cr-1Mo steel was proposed in this study. In this method, COD was calculated by classifying the components of COD; elastic, local plastic and fully plastic. In addition, the verification of this method was performed by comparing with the results of a series of four-point bending tests at elevated temperature using thin wall modified 9Cr-1Mo steel pipe containing a circumferential through wall crack. The rational leak rate evaluation method in LBB assessment was proposed.

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