Please wait a minute...
Frontiers of Mechanical Engineering

ISSN 2095-0233

ISSN 2095-0241(Online)

CN 11-5984/TH

Postal Subscription Code 80-975

2018 Impact Factor: 0.989

Front. Mech. Eng.    2018, Vol. 13 Issue (4) : 563-570    https://doi.org/10.1007/s11465-018-0487-9
RESEARCH ARTICLE
Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses
Jinya KATSUYAMA1(), Shumpei UNO2, Tadashi WATANABE3, Yinsheng LI1
1. Japan Atomic Energy Agency, Ibaraki 319-1195, Japan
2. Mizuho Information & Research Institute, Inc., Tokyo 101-8443, Japan
3. Fukui University, Fukui 914-0055, Japan
 Download: PDF(438 KB)   HTML
 Export: BibTeX | EndNote | Reference Manager | ProCite | RefWorks
Abstract

The thermal hydraulic (TH) behavior of coo-lant 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     
Corresponding Author(s): Jinya KATSUYAMA   
Just Accepted Date: 15 January 2018   Online First Date: 15 March 2018    Issue Date: 31 July 2018
 Cite this article:   
Jinya KATSUYAMA,Shumpei UNO,Tadashi WATANABE, et al. Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses[J]. Front. Mech. Eng., 2018, 13(4): 563-570.
 URL:  
https://academic.hep.com.cn/fme/EN/10.1007/s11465-018-0487-9
https://academic.hep.com.cn/fme/EN/Y2018/V13/I4/563
Fig.1  Nodalization diagram of a typical Japanese three-loop PWR plant model. (a) Japanese plant model nodalization; (b) primary side nodalization; (c) partially enlarged view of a primary Loop B; (d) secondary side nodalization of Loop B
Event a) Event time/s
Break 0
Reactor coolant pump trip 0
High head safety injection flow initiated due to ECCS (emergency core cooling system) operation 0
Main feedwater system stopped 0
Auxiliary feedwater started 0
Operator action to stop ECCS (Japanese) 600
Operator action to stop ECCS (the US) 1800
Tab.1  Event sequence of MSLB-103 transient [3]
Fig.2  Break flow rate during MSLB-103 transient
Fig.3  Inner pressure in primary system during MSLB-103 transient
Fig.4  Fluid temperature in downcomer during MSLB-103 transient
Fig.5  Water level in the pressurizer
Fig.6  Primary pressure obtained from downcomer node
Fig.7  Fluid temperature in downcomer
Fig.8  Heat transfer coefficient on RPV wall
Fig.9  FEA model for structural analyses. (a) Schematic image of RPV; (b) mesh division; (c) inner surface
Fig.10  Temperature distributions through the RPV wall obtained from the US and Japanese cases. (a) US case; (b) Japanese case
Fig.11  Time variation of hoop stress at the deepest point in postulated crack
1 Japan Electric Association. Verification Method of Fracture Toughness for In-service Reactor Pressure Vessel. JEAC4206-2016. 2016 (in Japanese)
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. 2006
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. 2004
4 Information Systems Laboratories. RELAP5/MOD3.3 CODE MANUAL. NUREG/CR-5535/Rev 1. 2003
5 Nuclear Safety Commission. Guidelines on the safety assessment of light water reactor type nuclear facilities. 1990 (in Japanese)
6 Dassault Systemes Simulia Corp. Abaqus analysis user’s manual. Version 6.14. 2015
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: 1439–1443
8 Japan Society of Mechanical Engineers. Steam Tables. 1999 (in Japanese)
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: 051402-1–051402-9
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, PVP2016-63433
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)
[1] Jidong KANG, James A. GIANETTO, William R. TYSON. Recent development in low-constraint fracture toughness testing for structural integrity assessment of pipelines[J]. Front. Mech. Eng., 2018, 13(4): 546-553.
[2] TU Shantung. Emerging challenges to structural integrity technology for high-temperature applications[J]. Front. Mech. Eng., 2007, 2(4): 375-387.
Viewed
Full text


Abstract

Cited

  Shared   
  Discussed