Hostname: page-component-cd9895bd7-lnqnp Total loading time: 0 Render date: 2024-12-29T07:45:54.091Z Has data issue: false hasContentIssue false
  • English
  • Français

A Model to Predict Stress Corrosion Cracking of Welded Stainless Steel Canister

Published online by Cambridge University Press:  20 February 2017

Gen Nakayama ,
Yohei Sakakibara ,
Tomomi Kouketsu ,
Kouji Arakawa ,
Yutaka Mizo ,
Iku Miyasaka ,
Yumiko Iwata ,
Susumu Kawakami ,
Hirotomo Setaka  and
Naoki Hieda
Gen Nakayama*
Affiliation:
Research Laboratory, IHI Corporation, 1, Shin-nakahara-cho, Isogo-ku, Yokohama 235-8501, Japan
Yohei Sakakibara
Affiliation:
Research Laboratory, IHI Corporation, 1, Shin-nakahara-cho, Isogo-ku, Yokohama 235-8501, Japan
Tomomi Kouketsu
Affiliation:
Research Laboratory, IHI Corporation, 1, Shin-nakahara-cho, Isogo-ku, Yokohama 235-8501, Japan
Kouji Arakawa
Affiliation:
Research Laboratory, IHI Corporation, 1, Shin-nakahara-cho, Isogo-ku, Yokohama 235-8501, Japan
Yutaka Mizo
Affiliation:
Production Engineering Center, IHI Corporation
Iku Miyasaka
Affiliation:
Nuclear Power Operations, IHI Corporation
Yumiko Iwata
Affiliation:
Nuclear Power Operations, IHI Corporation
Susumu Kawakami
Affiliation:
Nuclear Power Operations, IHI Corporation
Hirotomo Setaka
Affiliation:
Plant Project Department, IHI Corporation
Naoki Hieda
Affiliation:
Kitakyushu Liquefied Natural Gas CO., Inc., Kitakyuushu, 804-0002, Japan
*
*(Email: gen_nakayama@ihi.co.jp)
Get access

Abstract

Fabricated stainless steel structures are susceptible to stress corrosion cracking (SCC), despite being placed in chloride-containing natural water or humid atmospheres. The present paper describes a model that can define the conditions under which SCC is initiated and propagated, based on analyses of actual SCC incidents induced at welded flanges of cylindrical stainless steel structures.

Whenever the vitrified radioactive waste canister storage conditions deviate from normal and appropriate conditions due to earthquakes or tsunamis, the exposed canisters are expected to suffer SCC within 400 hours to 7 years, according to the analytical results obtained such as degree of sensitization, residual stress distribution, chloride ion concentration, and temperature.

Keywords

Corrosionsteelwaste management
Type
Articles
Information
MRS Advances , Volume 1 , Issue 63-64: Scientific Basis for Nuclear Waste Management XXXIX , 2016 , pp. 4179 - 4184
Copyright
Copyright © Materials Research Society 2017 

Access options

Get access to the full version of this content by using one of the access options below. (Log in options will check for institutional or personal access. Content may require purchase if you do not have access.)

References

REFERENCES

Nakayama, G.: The Piping Engineering (JPN), 2011.12, 33, Nihon-kogyo Publishing Co. (2011).Google Scholar
Nakayama, G. and Akashi, M., Zairyo-to-Kankyo, 48, 162, Japan Society of Corrosion Engineering (1999).CrossRefGoogle Scholar
Nakayama, G., Sakakibara, Y., ECS Trans. 2013 50(31), 303, Electrochemical Society (2013).Google Scholar
Nakayama, G. and Sakakibara, Y., Proc. of EUROCORR 2013, No. 1066, EFC (2013).Google Scholar
Bouch, P. J., et al., Inst. J of Pressure Vessels and Piping, 82, 299 (2005).Google Scholar
Kowaka, M., Kudo, T., J. Inst. Metal, Japan, 43, 595 (1979).Google Scholar
Chou, H., Mihara, K., Furukawa, K., Pseud SCC Defect Detection Methods Committee Report (JPN), Japan Non-destructive Inspection Society (2014).Google Scholar
RESOURCES and ENERGY AGENCY Japan High-level Radioactive Waste Materials and Formation Disposal (accessed 2016.3): http//www.enecho.meti.go.jp/category/electricityand.gas/nuclear/rw/hlw/qa/syo/syo03.html Google Scholar