Nuclear Engineering and Technology

  • 제30권5호
  • /
  • Pages.470-484
  • /
  • 1998
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지

Deterministic Fracture Mechanics Analysis of Pressurized Thermal Shock

  • M. J. Jhung (Korea Institute of Nuclear Safety) ;
  • Park, Y. W. (Korea Institute of Nuclear Safety)
  • 발행 : 1998.10.01
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

An analysis program for the evaluation of pressure vessel integrity under pressurized thermal shock (PTS) is developed. For given material properties and transient history such as temperature and pressure, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Using this program a round robin problem of PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study. The allowable maximum reference nil-ductility transition temperatures are determined for various crack sizes.

키워드

참고문헌

  1. Heat Conduction Oezisik, M.N.
  2. Analytical Method in Conduction Heat Transfer Myers, M.N.
  3. ASME Boiler and Pressure Vessel Code, Section Ⅲ, Appendix I ASME
  4. Theory of Elasticity(3rd ed.) Timoshenko, S.P.;Goodier, J.M.
  5. Theory and Design of Modern Pressure Vessels(2nd ed.) Harvey, J.F.
  6. ASME Boiler and Pressure Vessel Code, Section XI, Appendix A ASME
  7. Flaw Evaluation Procedures : ASME Section XI, EPRI NP-719-SR, EPRI Marston, T.U.
  8. Regulatory Guide 1.99, Rev.2 Radiation Embrittlement of Reactor Vessel Materials USNRC
  9. Nuclear Engineering and Design v.51 Application of Warm Prestressing Effects to Fracture Mechanics Analyses of Nuclear Reactor Vessels during Severe Thermal Shock McGowan, J.J.
  10. International Journal of Fracture v.22 A Model for Predicting the Influence of Warm Pre-stressing and Strain Ageing on the Cleavage Fracture Toughness of Ferritic Steels Curry, D.A.
  11. OECD/NEA PWG-3, Reactor Pressure Vessel Pressurized Thermal Shock International Comparative Assessment Study GRS mbH