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Integrity Evaluation of Korean Nuclear Reactor Pressure Vessel Under Pressurized Thermal Shock Conditions According to JEAC

[+] Author Affiliations
Se-Chang Kim, Jae-Boong Choi, Doo-Ho Cho

Sungkyunkwan University, Suwon, Gyeonggi, Korea

Sang-Min Lee, Yong-Beum Kim, Hae-Dong Chung

Korea Institute of Nuclear Safety, Daejeon, Korea

Paper No. PVP2013-97687, pp. V01BT01A057; 7 pages
doi:10.1115/PVP2013-97687
From:
  • ASME 2013 Pressure Vessels and Piping Conference
  • Volume 1B: Codes and Standards
  • Paris, France, July 14–18, 2013
  • Conference Sponsors: Pressure Vessels and Piping Division, Nondestructive Evaluation Engineering Division
  • ISBN: 978-0-7918-5564-5
  • Copyright © 2013 by ASME

abstract

In nuclear power plant, reactor pressure vessel (RPV) is the primary equipment that contains reactor cores and coolant. The RPV integrity should be evaluated in consideration with transient operation conditions and material deterioration. Especially, the pressurized thermal shock (PTS) has been considered as one of the most important issues regarding the RPV integrity since Rancho Seco nuclear power plant accident in1978.

In this paper, integrity evaluation of Korean RPV was performed by using finite element analysis. PTS conditions like small break loss of coolant accident (SBLOCA) and Turkey Point steam line break (TP-SLB) were applied as loading conditions. Neutron fluence data of actual RPV operated over 30 years was used to determine fracture toughness of RPV material.

The 3-dimensional finite element model including circumferential surface crack was generated for fracture mechanics analysis. The RPV integrity was evaluated according to Japan Electric Association Code (JEAC).

Copyright © 2013 by ASME

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