Integrity evaluation of Kori 1 reactor vessel for Rancho Seco transient

Rancho Seco Transient에 대한 고리 1호기 원자로용기의 건전성 평가

Jhung, M.J;Park, Y.W;Lee, J.B

  • Published : 1997.07.01


In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.


Pressurized Thermal Shock;Reactor Vessel;Structural Integrity;Critical Crack Depth;Effective Full Power Year;Crack Initiation;Crack Arrest