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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.
Yamashita, Takuya; Nagae, Yuji; Kikuchi, Koichi*; Yamamoto, Kenji*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Wakai, Takashi; Onizawa, Takashi; Kato, Takehiko*; Date, Shingo*; Kikuchi, Koichi*; Sato, Kenichiro*
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 8 Pages, 2013/07
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.
Ando, Masanori; Isobe, Nobuhiro*; Kikuchi, Koichi*; Enuma, Yasuhiro*
Nuclear Engineering and Design, 247, p.66 - 75, 2012/06
Times Cited Count：7 Percentile：48.31(Nuclear Science & Technology)The effect of ratcheting deformation on fatigue and creep-fatigue life in Mod.9Cr-1Mo steel was investigated. Uniaxial fatigue and creep-fatigue testing with superimposed strain were performed to evaluate the effect of ratcheting deformation on the failure cycle. In the fatigue tests with superimposed strain at 550
C, slight reductions of failure lives were observed. All of the numbers of cycles to failure in the fatigue tests with superimposed strain were within a factor of 1.5 of that of the fatigue test without superimposed strain at 550C. The apparent relationship between failure cycles and testing parameters was not observed. It was assumed that suppression of mean stress generation by cyclic softening reduces the effect of ratcheting strain. In the creep-fatigue tests with superimposed strain, test results indicated that the accumulated stain was negligible.
Ando, Masanori; Isobe, Nobuhiro*; Date, Shingo*; Kikuchi, Koichi*; Enuma, Yasuhiro*
Zairyo, 61(4), p.377 - 384, 2012/04
The effect of ratcheting deformation and pre-strain on fatigue and creep-fatigue life in Mod.9Cr-1Mo steel was investigated. Uni-axial fatigue and creep-fatigue tests with pre-strain and progressive strain were performed to evaluate the effect of pre-strain and ratcheting strain on the failure cycle. In the fatigue tests with pre-strain, failure lives were not declined. In the fatigue tests with progressive strain, slight reductions of failure lives were observed, however, they were within a factor of 1.5 of the failure life in normal fatigue test. In the creep-fatigue tests with progressive strain, the same conclusion of the fatigue tests was obtained. In both kinds of tests, maximum mean stresses during the tests were insignificant and/or generated in early cycle in the tests, and this character is considered as a reason of that the effect of ratcheting deformation on the fatigue and creep-fatigue lives are insignificant.
Nakamura, Kinya*; Ogata, Takanari*; Kikuchi, Hironobu; Iwai, Takashi; Nakajima, Kunihisa; Kato, Tetsuya*; Arai, Yasuo; Uozumi, Koichi*; Hijikata, Takatoshi*; Koyama, Tadafumi*; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 10(4), p.245 - 256, 2011/12
Sodium-bonded metallic fuel elements were fabricated for the first time in Japan for the irradiation test in the experimental fast test reactor JOYO. U-20Pu-10Zr fuel slugs of 200 mm in length and approximately 5 mm in diameter were fabricated in a small-scale injection casting furnace. Each fuel slug was loaded into the ferritic martenstic stainless steel (PNC-FMS) cladding tube with the sodium thermal bond, thermal insulator and reflector in a helium gas atmosphere glove box. After top-end plug welding to the cladding tube and heat treatment of the welding area, each fuel element was subjected to the sodium bonding process. After the inspection such as element length, gas plenum length and helium-leak tightness, six metallic fuel elements are transported to the JOYO site for the coming irradiation test.
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
Deguchi, Yoshihiro*; Imanaka, Koichi*; Takata, Takashi*; Yamaguchi, Akira*; Kikuchi, Shin; Ohshima, Hiroyuki
Proceedings of 3rd Asian Symposium on Computational Heat Transfer and Fluid Flow (ASCHT 2011) (CD-ROM), 6 Pages, 2011/09
In a sodium-cooled fast reactor (SFR), liquid sodium is used as a heat transfer fluid because of its excellent heat transport capability. On the other hand, it has strong chemical reactivity with water vapor. One of the design basis accidents of the SFR is the water leakage into the liquid sodium flow by a breach of heat transfer tubes. This process ends up damages on the heat transport equipment in the SFR. Therefore, the study on sodium-water chemical reactions is of paramount importance for security reasons. This study aims to clarify the sodium-water reaction mechanisms using an elementary reaction analysis. As a result of the analysis, It was demonstrated that the main reaction is Na+H
O 
NaOH+H as gas-phase reaction.
Wakai, Takashi; Machida, Hideo*; Yoshida, Shinji*; Kawashima, Fumiko*; Kikuchi, Koichi*; Xu, Y.*; Tsukimori, Kazuyuki
Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 10 Pages, 2011/07
This paper describes the improved COD assessment method and verifies the validity of the method based on 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. As a result, COD values calculated by the proposed method were in a good agreement with the experimental results for the uniform pipe without weld. In the case that the crack was machined at weld metal or heat affected zone (HAZ), proposed method predicted relatively larger COD than the experimental results. The causes of such discrepancies were discussed comparing with the results of finite element analyses. Based on these examinations, the rational leak rate evaluation method in LBB assessment was proposed.
Wakai, Takashi; Machida, Hideo*; Yoshida, Shinji*; Tokiyoshi, Takumi*; Kikuchi, Koichi*; Xu, Y.*; Tsukimori, Kazuyuki
Proceedings of 2011 ASME Pressure Vessels and Piping Conference (PVP 2011) (CD-ROM), 8 Pages, 2011/07
This paper studies the master curve to estimate the crack length when a postulated initial crack unexpectedly grows and penetrates the pipe thickness. In order to obtain the fatigue crack and creep crack growth characteristics of modified 9Cr-1Mo steel pipes, fatigue crack and creep crack growth tests were conducted using compact tension (CT) specimens and crack growth rates for both fatigue and creep at elevated temperature were obtained. Based on the obtained material characteristics and the results of a series of crack growth calculations, a relationship between the penetrated crack length and the ratio of membrane to total stress, so called as master curve, was proposed. In this study, master curves were proposed for pipes made of modified 9Cr-1Mo steel as a function of pipe geometry, i.e. the ratio of radius to thickness.
Wakai, Takashi; Nagae, Yuji; Takaya, Shigeru; Obara, Satoshi; Date, Shingo*; Yamamoto, Kenji*; Kikuchi, Koichi*; Sato, Kenichiro*
Tainetsu Kinzoku Zairyo Dai-123-Iinkai Kenkyu Hokoku, 52(2), p.147 - 159, 2011/07
By employing high-Cr ferritic steels to the structural materials for JSFR, a compact plant designing can be achieved. It contributes to reduce the construction cost and to enhance the freedom of designing. Among the high-Cr ferritic steels, modified 9Cr-1Mo steel (compatible to ASTM A335 P91) is a candidate of the structural material for the demonstration facility of JSFR, because the steel has superior microstructure stability at elevated temperature for long time. However, remarkable creep strength degradation has been observed in the welded joint of high-Cr ferritic steels, especially in long-term region. It is known as "Type-IV damage". In the elevated temperature designing for the fast reactors, such creep strength degradation must be taken into account properly. Therefore, the creep strength assessment procedure and the allowable stress for the welded joints made of modified 9Cr-1Mo steel have been proposed. In this study, (1) a series of creep rupture tests to verify the validity of the creep strength assessment procedure was performed. (2) Applicability of the creep strength assessment procedure to the creep fatigue strength assessment of the welded joints made of modified 9Cr-1Mo steel was investigated. (3) Metallurgical examinations of creep ruptured specimens were carried out to confirm the contribution of "Type-IV damage". As a result, it was demonstrated that the creep strength assessment procedure was validated using the long-term creep rupture test results less than 30,000h and that the creep strength assessment procedure was applicable to the creep-fatigue strength assessment based on some uniaxial creep-fatigue test results.
Ando, Masanori; Isobe, Nobuhiro*; Date, Shingo*; Kikuchi, Koichi*; Enuma, Yasuhiro*
Dai-48-Kai Koon Kyodo Shimpojiumu Maezurishu, p.110 - 114, 2010/12
no abstracts in English
Kawasaki, Nobuchika; Date, Shingo*; Kikuchi, Koichi*; Isobe, Nobuhiro*; Kasahara, Naoto
Nihon Kikai Gakkai M&M 2009 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.535 - 536, 2009/07
The effect of ratcheting strain on creep-fatigue strength was investigated in order to adjust the strain limit. Ratcheting creep-fatigue tests were conducted at 600 C with 1 hr hold time. Creep fatigue lives did not decrease when the accumulated strain was superimposed within 5%. Therefore the effect of ratcheting on creep-fatigue strength is negligible when accumulated strain exceeds 2%.
Okajima, Satoshi; Date, Shingo*; Kawasaki, Nobuchika; Kikuchi, Koichi*; Isobe, Nobuhiro*; Kasahara, Naoto
Nihon Kikai Gakkai M&M 2009 Zairyo Rikigaku Kanfarensu Koen Rombunshu (CD-ROM), p.537 - 538, 2009/07
no abstracts in English
Nakazawa, Takashi; Kikuchi, Hisaki; Yasu, Katsuji; Yoshino, Toshiaki; Ashikagaya, Yoshinobu; Sato, Koichi; Minowa, Yuji; Nomura, Toshibumi
JAERI-Tech 2001-010, 125 Pages, 2001/03
no abstracts in English
; Nakazawa, Takashi; Sato, Koichi; Kikuchi, Hisaki;
KURRI-KR-30, p.42 - 47, 1998/00
no abstracts in English
Kawasaki, Nobuchika; Kato, Shoichi; Yamauchi, Masafumi*; Nagae, Yuji; Kikuchi, Koichi*; Kasahara, Naoto
no journal, ,
To apply commercialized fast reactor, JSFR, creep fatigue evaluation method is developing to consider holding position during creep. In general thermal transients, creep hold position is intermediate position in the strain range. To consider holding position in the evaluation method, uni-axial intermediate hold type creep fatigue tests were performed and verified the evaluation method.
Nakamura, Kinya*; Kikuchi, Hironobu; Ogata, Takanari*; Iwai, Takashi; Arai, Yasuo; Uozumi, Koichi*; Hijikata, Takatoshi*; Koyama, Tadafumi*; Itagaki, Wataru; Soga, Tomonori
no journal, ,
Irradiation test of metallic fuel at fast test reactor JOYO is planned. By using techniques of casting, fabrication, analysis and inspection, metallic fuel elements were made for the first time in Japan. In this paper, fabrication technique of metallic fuel elements is reported. After filling up thermal bond (Na), a reflector element, a thermal shelter element and a fuel slug, the upper plug was welded. Sodium bonding treatment was carried out by vibrating the fuel element under heating. By inspection, it was confirmed that the specifications were satisfied.
Deguchi, Yoshihiro*; Imanaka, Koichi*; Takata, Takashi*; Yamaguchi, Akira*; Kikuchi, Shin; Ohshima, Hiroyuki
no journal, ,
If the heat transfer tube in the steam generator of a sodium-cooled fast reactor is failed, high pressurized water vapor blows into the liquid sodium and the sodium-water reaction (SWR) takes place. Japan Atomic Energy Agency has been developing a multi-dimensional sodium-water reaction code for the analytical evaluation of this reaction. One of the items of code development is to construct the applicable chemical model for SWR. In this study, elementary reaction model has been developed for the purpose of identifying the dominant overall reactions for SWR. Also, we applied this model to the sodium-water counter-flow reaction field and evaluated the major reaction pathways.
