About: Irradiation assisted stress corrosion cracking of austenitic stainless steel WWER reactor core internals     Goto   Sponge   NotDistinct   Permalink

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  • The neutron irradiation changes the material's microstructure and mechanical properties, basis of occurrence and increased sensitivity to IASCC. This paper surveys new results regarding IASCC of irradiated austenitic Ti-stabilized stainless steel 08Ch18N10T from WWER 440 Greifswald decommissioned after 15 years in service. Three components LWR irradiated to 2-19 dpa were tested. IASCC was investigated by Slow Strain Rate and Crack Growth Rate tests in simulated water 320°C. IASCC was judged according IG+TG fracture occurrence. Without irradiation components do not suffer SCC in the water. However, areas of mixed IG and TG fracture appeared on specimens. Tests represent different stress strain conditions for IASCC initiation and growth. Effects of SSRT strain rate and CGR test load level were found to be significant for IASCC. Total IG+TG fraction of SSRT ranged 1 to 18% and decreased with decreasing strain rate. Results are compared with previous data on the fast reactor irradiated material.
  • The neutron irradiation changes the material's microstructure and mechanical properties, basis of occurrence and increased sensitivity to IASCC. This paper surveys new results regarding IASCC of irradiated austenitic Ti-stabilized stainless steel 08Ch18N10T from WWER 440 Greifswald decommissioned after 15 years in service. Three components LWR irradiated to 2-19 dpa were tested. IASCC was investigated by Slow Strain Rate and Crack Growth Rate tests in simulated water 320°C. IASCC was judged according IG+TG fracture occurrence. Without irradiation components do not suffer SCC in the water. However, areas of mixed IG and TG fracture appeared on specimens. Tests represent different stress strain conditions for IASCC initiation and growth. Effects of SSRT strain rate and CGR test load level were found to be significant for IASCC. Total IG+TG fraction of SSRT ranged 1 to 18% and decreased with decreasing strain rate. Results are compared with previous data on the fast reactor irradiated material. (en)
Title
  • Irradiation assisted stress corrosion cracking of austenitic stainless steel WWER reactor core internals
  • Irradiation assisted stress corrosion cracking of austenitic stainless steel WWER reactor core internals (en)
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  • Irradiation assisted stress corrosion cracking of austenitic stainless steel WWER reactor core internals
  • Irradiation assisted stress corrosion cracking of austenitic stainless steel WWER reactor core internals (en)
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  • RIV/46356088:_____/11:#0001128!RIV12-MPO-46356088
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  • 205850
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  • RIV/46356088:_____/11:#0001128
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  • corrosion; cracking; irradiation; materials structure (en)
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  • [CC207D2BF235]
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  • Colorado Springs, USA
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  • USA
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  • Environmental Degradation of Materials in Nuclear Power Systems: Water Reactors (Proceedings of the 15th International Conference)
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  • Hojná, Anna
  • Ernestová, Miroslava
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number of pages
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  • Wiley
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  • 978-1-118-13241-8
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