Advertisement

Frontiers of Mechanical Engineering

, Volume 13, Issue 4, pp 563–570 | Cite as

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

  • Jinya Katsuyama
  • Shumpei Uno
  • Tadashi Watanabe
  • Yinsheng Li
Research Article
  • 39 Downloads
Part of the following topical collections:
  1. High Parameter Pressure Equipment

Abstract

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

Keywords

structural integrity reactor pressure vessel pressurized thermal shock thermal hydraulic analysis pressurized water reactor weld residual stress 

Preview

Unable to display preview. Download preview PDF.

Unable to display preview. Download preview PDF.

Notes

Acknowledgements

This study was performed under the contract research entrusted from the Secretariat of Nuclear Regulation Authority of Japan.

References

  1. 1.
    Japan Electric Association. Verification Method of Fracture Toughness for In-service Reactor Pressure Vessel. JEAC4206-2016. 2016 (in Japanese)Google Scholar
  2. 2.
    U.S. Nuclear Regulatory Commission. Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10CFR50.61) Summary Report. NUREG-1806. 2006Google Scholar
  3. 3.
    U.S. Nuclear Regulatory Commission. RELAP5 Thermal Hydraulic Analysis to Support PTS Evaluations for the Oconee-1, Beaver Valley-1, and Palisades Nuclear Power Plants. NUREG/CR-6858. 2004Google Scholar
  4. 4.
    Information Systems Laboratories. RELAP5/MOD3.3 CODE MANUAL. NUREG/CR-5535/Rev 1. 2003Google Scholar
  5. 5.
    Nuclear Safety Commission. Guidelines on the safety assessment of light water reactor type nuclear facilities. 1990 (in Japanese)Google Scholar
  6. 6.
    Dassault Systemes Simulia Corp. Abaqus analysis user’s manual. Version 6.14. 2015Google Scholar
  7. 7.
    Watanabe T. Effects of ECCS on the cold-leg fluid temperature during SGTR accidents. International Journal of Mechanical, Aerospace, Industrial, Mechatronic and Manufacturing Engineering, 2015, 9: 1439–1443Google Scholar
  8. 8.
    Japan Society of Mechanical Engineers. Steam Tables. 1999 (in Japanese)Google Scholar
  9. 9.
    Katsuyama J, Nishikawa H, Udagawa M, et al. Assessment of residual stress due to overlay-welded cladding and structural integrity of a reactor pressure vessel. Journal of Pressure Vessel Technology, 2013, 135: 051402-1–051402-9CrossRefGoogle Scholar
  10. 10.
    Uno S, Katsuyama J, Watanabe T, et al. Loading condition evaluation for structural integrity assessment of RPV due to PTS event based on three-dimensional thermal-hydraulics and structural analyses. In: Proceedings of the ASME 2016 Pressure Vessels & Piping Conference. 2016, PVP2016-63433Google Scholar
  11. 11.
    Japan Society of Mechanical Engineers. Codes for Nuclear Power Generation Facilities—Rules on Fitness-for-Service for Nuclear Power Plants. JSME S NA1-2012, 2012 (in Japanese)Google Scholar

Copyright information

© Higher Education Press and Springer-Verlag GmbH Germany, part of Springer Nature 2018

Authors and Affiliations

  • Jinya Katsuyama
    • 1
  • Shumpei Uno
    • 2
  • Tadashi Watanabe
    • 3
  • Yinsheng Li
    • 1
  1. 1.Japan Atomic Energy AgencyIbarakiJapan
  2. 2.Mizuho Information & Research Institute, Inc.TokyoJapan
  3. 3.Fukui UniversityFukuiJapan

Personalised recommendations