0
Research Papers

Application of the System Based Code Concept to the Determination of In-Service Inspection Requirements

[+] Author and Article Information
Shigeru Takaya

Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393, Japan e-mail: takaya.shigeru@jaea.go.jp

Tai Asayama

Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393, Japan e-mail: asayama.tai@jaea.go.jp

Yoshio Kamishima

Mitsubishi FBR systems, Inc., 2-34-17, Jingumae Shibuya-ku, Tokyo 150-0001, Japan e-mail: yoshio_kamishima@mfbr.mhi.co.jp

Hideo Machida

TEPCO Systems Corporation, 2-37-28, Eitai Koto-ku, Tokyo 135-0034, Japan e-mail: machida-hideo@tepsys.co.jp

Daigo Watanabe

Mitsubishi Heavy Industries, Ltd., 5-717-1 Fukahori, Nagasaki, Nagasaki 851-0392, Japan e-mail: daigo_watanabe@mhi.co.jp

Satoru Nakai

Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393, Japan e-mail: nakai.satoru@jaea.go.jp

Masaki Morishita

Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393, Japan e-mail: morishita.masaki@jaea.go.jp

1Corresponding author.

Manuscript received July 16, 2014; final manuscript received September 29, 2014; published online February 9, 2015. Assoc. Editor: Joseph Miller.

ASME J of Nuclear Rad Sci 1(1), 011004 (Feb 09, 2015) (9 pages) Paper No: NERS-14-1020; doi: 10.1115/1.4026392 History: Received July 16, 2014; Accepted November 14, 2014; Online February 09, 2015

A new process for determination of in-service inspection (ISI) requirements was proposed on the basis of 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.

FIGURES IN THIS ARTICLE
<>
Copyright © 2015 by ASME
Your Session has timed out. Please sign back in to continue.

References

ASME, 2013, “Boiler and Pressure Vessel Code, Section XI, Division 1, Rules for Inspection and Testing of Components of Light-Water-Cooled Plants.”
ASME, 2001, “Boiler and Pressure Vessel Code, Section XI, Division 3, Rules for Inspection and Testing of Components of Liquid-Metal-Cooled Plants.”
Rao K. R., ed., 2006, Companion Guide to the ASME Boiler & Pressure Vessel Code, 2nd ed., Vol. 2, ASME Press, New York.
Asada, Y., Tashimo, M., and Ueta, M., 2002, “System Based Code—Principal Concept,” Proceedings of the 10th International Conference on Nuclear Engineering, Paper No. 22730.
Asada, Y., Tashimo, M., and Ueta, M., 2002, “System Based Code—Basic Structure,” Proceedings of the 10th International Conference on Nuclear Engineering, Paper No. 22731.
Asada, Y., 2006, “Japanese Activities Concerning Nuclear Codes and Standards—Part II,” J. Press. Vess. Technol., 128(1), pp. 64–70. 10.1115/1.2138063
Power Reactor and Nuclear Fuel Development Corporation (Former Japan Atomic Energy Agency), 1981, “Elevated Temperature Structural Design Guide for Class 1 Components of Prototype Fast Reactors,” (in Japanese).
JSME, 2013, “Codes for Nuclear Power Generation Facilities, Rules on Design and Construction for Nuclear Power Plants, Part II: Fast Reactor Standards,” (in Japanese).
ASME, 2013, “Boiler and Pressure Vessel Code, Section XI, Division 1, Nonmandatory Appendix R,” Risk-Informed Inspection Requirements for Piping.
AESJ, 2008, “Code on Implementation and Review of Nuclear Power Plant Ageing Management Programs,” .
Power Reactor and Nuclear Fuel Development Corporation (Former Japan Atomic Energy Agency), 1997, “Summary of Implementation Status of the Comprehensive Review of the Monju Design and Operational Safety,” http://www.jaea.go.jp/jnc/pnc-news/ntopic/PT97/P9711/PE97112702/honbun/index.html (in Japanese).
ASME, 2013, “Boiler and Pressure Vessel Code, Section XI, Division 1, Nonmandatory Appendix N,” Flow-Induced Vibrations of Tubes and Tube Banks.
Power Reactor and Nuclear Fuel Development Corporation (Former Japan Atomic Energy Agency), 1984, “Interpretations: Elevated Temperature Structural Design Guide for Fast Breeder Reactors, Material Strength Standards,” (in Japanese).
Central Research Institute of Electric Power Industry, 1994, “The Draft of the Guideline for Application of Inelastic Fracture Mechanics to Fast Reactor Components” (in Japanese).
Odaka, S., Kato, S., Kawakami, T., Suzuki, T., and Takamori, Y., 2003, “Material test data of SUS304 Steel (III),” (in Japanese).
Kurisaka, K., Nakai, R., Asayama, T., and Takaya, S., 2011, “Development of System Based Code (1) Reliability Target Derivation of Structures and Components,” J. Power Energy Sys., 5, pp. 19–32. [CrossRef]
Yamano, H., Kubo, S., Shimakawa, Y., Fujita, K., Suzuki, T., and Kurisaka, K., 2012, “Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor,” Sci. Technol. Nucl. Install., Vol. 2012, ID 614973, doi: .

Figures

Grahic Jump Location
Fig. 1

Stage I evaluation of the proposed ISI requirement determination process

Grahic Jump Location
Fig. 2

Stage II evaluation of the proposed ISI requirement determination process

Grahic Jump Location
Fig. 3

Monju reactor main components

Grahic Jump Location
Fig. 4

Dimensions of the UCS in the vicinity of sodium surface level

Grahic Jump Location
Fig. 5

Probability distribution of crack depth

Tables

Errata

Discussions

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In