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). For sodium pipes of JSFR, 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 low. 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 COD, an appropriate COD assessment method must be established to perform LBB assessment. However, COD assessment method applicable for JSFR sodium pipes—thin wall and small work hardening material—has not been proposed yet. Thus, 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. As a result, COD values calculated by the proposed method were in a good agreement with the experimental results for the uniform pipe without a weld. In the case that the crack was machined in the weld metal or heat affected zone (HAZ), the proposed method predicted relatively larger COD than the experimental results. The causes of such discrepancies were discussed by comparing with the results of finite element analyses. Based on these examinations, the rational leak rate evaluation method in LBB assessment was proposed.
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Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki 3111393,
e-mail: wakai.takashi@jaea.go.jp
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
e-mail: yoshida-shinji@tepsys.co.jp
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki, 3111393,
e-mail: enuma.yasuhiro@jaea.go.jp
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February 2013
Research-Article
Development of Leak-Before-Break Assessment Method for Japan Sodium Cooled Fast Reactor Pipe—Part 1 Crack Opening Displacement Assessment of Thin Wall Pipes Made of Modified 9Cr-1Mo Steel
Takashi Wakai,
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki 3111393,
e-mail: wakai.takashi@jaea.go.jp
Takashi Wakai
1
Principal Researcher
Structural Reliability Research Group,
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki 3111393,
Japan
e-mail: wakai.takashi@jaea.go.jp
1Presently Japan Atomic Energy Agency.
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Hideo Machida,
Hideo Machida
General Manager
e-mail: machida-hideo@tepsys.co.jp
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Manabu Arakawa,
Manabu Arakawa
Chief Engineer
e-mail: arakawa-manabu@tepsys.co.jp
TEPCO Systems Corporation,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
Nuclear Plant Engineering Department
,TEPCO Systems Corporation,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
Japan
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Shinji Yoshida,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
e-mail: yoshida-shinji@tepsys.co.jp
Shinji Yoshida
TEPCO Systems Corporation,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
Japan
e-mail: yoshida-shinji@tepsys.co.jp
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Yasuhiro Enuma
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki, 3111393,
e-mail: enuma.yasuhiro@jaea.go.jp
Yasuhiro Enuma
JSFR Systems Development Planning Office,
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki, 3111393,
Japan
e-mail: enuma.yasuhiro@jaea.go.jp
Search for other works by this author on:
Takashi Wakai
Principal Researcher
Structural Reliability Research Group,
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki 3111393,
Japan
e-mail: wakai.takashi@jaea.go.jp
Hideo Machida
General Manager
e-mail: machida-hideo@tepsys.co.jp
Manabu Arakawa
Chief Engineer
e-mail: arakawa-manabu@tepsys.co.jp
TEPCO Systems Corporation,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
Nuclear Plant Engineering Department
,TEPCO Systems Corporation,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
Japan
Shinji Yoshida
TEPCO Systems Corporation,
2-37-28 Eitai Koto-ku,
Tokyo 1350034,
Japan
e-mail: yoshida-shinji@tepsys.co.jp
Yasuhiro Enuma
JSFR Systems Development Planning Office,
Japan Atomic Energy Agency,
4002 Narita-cho Oarai,
Ibaraki, 3111393,
Japan
e-mail: enuma.yasuhiro@jaea.go.jp
1Presently Japan Atomic Energy Agency.
Contributed by the Pressure Vessel and Piping Division of ASME for publication in the Journal of Pressure Vessel Technology. Manuscript received May 9, 2011; final manuscript received June 15, 2012; published online December 4, 2012. Assoc. Editor: Kunio Hasegawa.
J. Pressure Vessel Technol. Feb 2013, 135(1): 011401 (9 pages)
Published Online: December 4, 2012
Article history
Received:
May 9, 2011
Revision Received:
June 15, 2012
Citation
Wakai, T., Machida, H., Arakawa, M., Yoshida, S., and Enuma, Y. (December 4, 2012). "Development of Leak-Before-Break Assessment Method for Japan Sodium Cooled Fast Reactor Pipe—Part 1 Crack Opening Displacement Assessment of Thin Wall Pipes Made of Modified 9Cr-1Mo Steel." ASME. J. Pressure Vessel Technol. February 2013; 135(1): 011401. https://doi.org/10.1115/1.4007642
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