Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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PWSCC growth rate model of alloy 690 for head penetration nozzles of Korean PWRs

  • Kim, Sung-Woo (Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Eom, Ki-Hyun (Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Lim, Yun-Soo (Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute) ;
  • Kim, Dong-Jin (Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute)
  • Received : 2018.09.17
  • Accepted : 2019.01.16
  • Published : 2019.05.25
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Abstract

This work aims to establish a model of a primary water stress corrosion crack growth rate of Alloy 690 material for the head penetration nozzles of Korean pressurized water reactors. The test material had an inhomogeneous microstructure with bands of fine-grains and intragranular carbides in the matrix of coarse-grains, which was similar to the archive materials of the head penetration nozzles. The crack growth rate was measured from the strain-hardened materials as a function of the stress intensity factor in simulated primary water at various temperatures and dissolved hydrogen contents. The effects of strain-hardening, temperature, and dissolved hydrogen on the crack growth rate were analyzed independently, and were then introduced as normalizing factors in the crack growth rate model. The crack growth rate model proposed in this work provides a key element of the tools needed to assess the progress of a stress corrosion crack when detected in thick-wall Alloy 690 components in Korean reactors.

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References

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