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Journal Articles

A Study for proposal of welded joint strength reduction factors of modified 9Cr-1Mo steel for Japan sodium cooled fast reactor (JSFR)

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

Journal Articles

Evaluation of excessive loading effect on fatigue crack growth behavior based on crack blunting and stress distribution in front of the crack tip

Yamaguchi, Yoshihito; Katsuyama, Jinya; Onizawa, Kunio; Li, Y.*

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

We performed fatigue crack growth tests under constant amplitude cyclic loading with a single excessive tensile/compressive load. The stress distribution in front of crack tip and crack blunting were estimated by FEM analyses. After the crack tip was blunted by the excessive tensile loading, the effect of the excessive loading on crack growth rate varied depending on the magnitude of the subsequent compressive loading. When a compressive load is enough to close the crack, the crack growth rate became higher than that before the excessive tensile loading while increasing the tensile stress in front of crack tip. A crack growth prediction method has been proposed considering the effects of the excessive loading based on the variation of the stress distribution in front of crack tip and the crack blunting. The predicted crack growth rate by the proposed method was correlated with the experimental ones.

Journal Articles

Alternative reference temperature based on master curve approach in Japanese reactor pressure vessel steels

Hirota, Takatoshi*; Hirano, Takashi*; Onizawa, Kunio

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

Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011 was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature To can be determined according to this code in Japan. In this study, using the existing fracture toughness data of Japanese RPV steels, the method for determination of the alternative reference temperature RT$$_{To}$$ based on Master Curve reference temperature To was statistically examined, so that RT$$_{To}$$ has an equivalent safety margin to the conventional RT$$_{NDT}$$. Through the statistical treatment, the alternative reference temperature RT$$_{To}$$ was proposed as the following equation; RT$$_{To}$$ = To + C$$_{MC}$$ + 2$$sigma$$$$_{To}$$. This method is applicable to the Japan Electric Association Code JEAC 4206 as an option item.

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

Fracture toughness evaluation of reactor pressure vessel steels by master curve method using Mini-CT specimens

Tobita, Toru; Nishiyama, Yutaka; Otsu, Takuyo; Udagawa, Makoto; Katsuyama, Jinya; Onizawa, Kunio

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

Mini-CT (0.16T-CT) specimens up to eight can be taken from broken halves of surveillance Charpy specimens. We conducted a series of fracture toughness tests based on the Master curve method for several specimen size and shapes, such as 0.16T-CT, pre-cracked Charpy type, 0.4T-CT and 1T-CT specimens, in commercially manufactured 5 kinds of A533B class1 steels with different impurity contents and fracture toughness levels. Reference temperature To of 0.16T-CT specimens was approximately equal to those of 1T-CT and other type of specimens for all materials. We also examined a loading rate effect on TO of Mini-CT specimens for some materials within the specified range in the test method. Higher loading rate gave rise to slightly higher TO. The difference in TO between upper and lower loading rate of the standard was approximately 10 degree C.

Journal Articles

A Round robin program of master curve evaluation using miniature C(T) specimens, 2; Fracture toughness comparison in specified loading rate condition

Yamamoto, Masato*; Onizawa, Kunio; Yoshimoto, Kentaro*; Ogawa, Takuya*; Mabuchi, Yasuhiro*; Miura, Naoki*

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

Master Curve (MC) approach for the fracture toughness evaluation is expected to be a powerful tool to assess the structural integrity of reactor pressure vessels (RPVs). In order to get sufficient number of reliable data by the MC approach from used specimens of surveillance tests for RPVs, the use of miniature specimens is necessary. For this purpose, a round robin test program on the miniature compact tension specimens (Mini-CT) of 4 mm thick for the MC approach was launched with the participation of academia, industries and a research institute in Japan. The program aims to verify the reliability of experimental data from Mini-CT, and to pick out technical issues to be solved. As the second step of this program, the effect of loading rate (d$$K$$/d$$t$$) was evaluated based on enlarged database. Despite of the difference in d$$K$$/d$$t$$, no specific difference in scatter band of $$T_{rm 0}$$ was found in d$$K$$/d$$t$$ - $$T_{rm 0}$$ relationships. D$$K$$/d$$t$$ seems not to be sensitive on scatter band of To in the present results.

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

Study on piping response under multiple excitation, 1; Triple shaking table test of piping having three-supporting points

Watakabe, Tomoyoshi; Kaneko, Naoaki*; Aida, Shigekazu*; Otani, Akihito*; Moriizumi, Makoto*; Tsukimori, Kazuyuki; Kitamura, Seiji

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

The piping in a nuclear power plant is laid across multiple floors of a single building or two buildings, which are supported at many points. As the piping is excited by multiple inputs from the supporting points during an earthquake, seismic response analysis by multiple excitations is needed to obtain the exact seismic response of the piping. However, few experiments involving such multiple excitations have been performed to verify the validity of multiple excitation analysis. To perform rational seismic design and evaluation, it is important to investigate the seismic response by multiple excitations and verify the validity of the analysis method by multiple excitation test. This paper reports on the result of the shaking test using triple uni-axial shaking tables and a 3-dimensional piping model.

Journal Articles

Study on piping response under multiple excitation, 2; Validation for multiple analysis of piping

Kai, Satoru*; Watakabe, Tomoyoshi; Kaneko, Naoaki*; Tochiki, Kunihiro*; Moriizumi, Makoto; Tsukimori, Kazuyuki

Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 9 Pages, 2013/07

The piping in a nuclear power plant is laid across multiple floors of a single building or two buildings, which are supported at many points. As the piping is excited by multiple inputs from the supporting points during an earthquake, seismic response analysis by multiple excitations is needed to obtain the exact seismic response of the piping. However, few experiments involving such multiple excitations have been performed to verify the validity of multiple excitation analysis. To perform rational seismic design and evaluation, it is important to investigate the seismic response by multiple excitations and verify the validity of the analysis method by multiple excitation test. This paper reports the validation result of the multiple excitation analysis of piping compared with the results of the multiple excitations shaking test by using triple uni-axial shaking table and a 3-dimensional piping model (89.1 mm diameter and 5.5 mm thickness).

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

Journal Articles

Development of 2012 edition of JSME code for design and construction of fast reactors, 2; Development of the material strength standard of 316FR stainless 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

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.

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