Corrosion Science and Technology

The Corrosion Science Society of Korea (한국부식방식학회)
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Failure Evaluation Plan of a Reactor Internal Components of a Decommissioned Plant

  • Received : 2021.07.02
  • Accepted : 2021.07.09
  • Published : 2021.08.31
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Abstract

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

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Acknowledgement

This work was supported by Korea Institute of Energy Technology Evaluation and Planning (KETEP) grant funded by the Korea government (MOTIE) (20191510301140, Development of Failure and Degradation Analysis Technologies of Reactor Internal Baffle Former Bolts from Decommissioning NPP).

References

  1. Z. Jiao, J. T. Busby, and G. S. Was, Deformation microstructure of proton-irradiated stainless steels, Journal of Nuclear Materials, 361, 218 (2007). Doi: https://doi.org/10.1016/j.jnucmat.2006.12.012
  2. O. K. Chopra and A. S. Rao, A review of irradiation effects on LWR core internal materials - IASCC susceptibility and crack growth rates of austenitic stainless steels, Journal of Nuclear Materials, 409, 235 (2011). Doi: https://doi.org/10.1016/j.jnucmat.2010.12.001
  3. Y. S. Lim, D. J. Kim, S. S.Hwang, M. J. Choi, and S. W. Cho, Effects of Proton Irradiation on the Microstructure and Surface Oxidation Characteristics of Type 316 Stainless Steel, Corrosion Science and Technology, 20, 158 (2021). Doi: https://doi.org/10.14773/cst.2021.20.3.158
  4. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), EPRI, Palo Alto, CA, USA (2019).
  5. S. S. Hwang, et al., PRIMA-NET workshop, Development of failure and degradation analysis technologies of reactor internal baffle former bolts from decommissioning NPP, Jan. 16, Gimcheon, Korea (2020).
  6. J. Smith, EPRI - NRC Materials Meeting, EPRI Irradiated Materials Testing Programs, ML13162A571, USNRC, USA (2013). https://www.nrc.gov/docs/ML1316/ML13162A571.pdf
  7. C. G Lee, Project for facility modification of baffle former bolt retrieval from Kori unit 1, KEPCO KPS (2020).
  8. ASTM E 8-01, Standard Test Methods for Tension Testing of Metallic Materials, ASTM International, West Conshohocken, PA (2001). Doi: https://doi.org/10.1520/E0008-01
  9. K. H. Na, et al., KNS workshop, Research progress and plan for the utilization of Kori 1 harvesting materials, Oct. 23, Ilsan, Korea (2019).
  10. S. Hwang, et al., Development of Life Prediction and Extension Technologies of Nuclear Reactor Internals, KAERI/RR-4098/2016, Daejeon, Korea (2016).
  11. G. F. Somville, Proc. 17th International conference on nuclear engineering (ICONE17), ASME, July 12-16, Brussels, Belgium (2009).