Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
권호 표지
DOI QR코드

DOI QR Code

FATIGUE ANALYSIS OF A REACTOR PRESSURE VESSEL FOR SMART

  • Jhung, Myung-Jo (Research Management Department, Korea Institute of Nuclear Safety)
  • Received : 2011.06.13
  • Accepted : 2011.07.16
  • Published : 2012.08.25
Download PDF
⟨ Previous Next ⟩

Abstract

The structural integrity of mechanical components during several transients should be assured in the design stage. This requires a fatigue analysis including thermal and stress analyses. As an example, this study performs a fatigue analysis of the reactor pressure vessel of SMART during arbitrary transients. Using heat transfer coefficients determined based on the operating environments, a transient thermal analysis is performed and the results are applied to a finite element model along with the pressure to calculate the stresses. The total stress intensity range and cumulative fatigue usage factor are investigated to determine the adequacy of the design.

Keywords

References

  1. IAEA, Status Report for Advanced Nuclear Reactor Designs - Report 77 - System-Integrated Modular Advanced Reactor (SMART), http://aris.iaea.org/ARIS/, 2011, International Atomic Energy Agency.
  2. ANSYS, Inc., Theory Reference for ANSYS and ANSYS Workbench Release 12.0, 2009, Canonsburg, PA.
  3. Holman, J.P., Heat Transfer, 9th ed., 2002, McGraw-Hill, London.
  4. Wagner, W., Kretzschmar, H-J., International Steam Tables - Properties of Water and Steam based on the Industrial Formulation IAPWS-IF97, 2nd ed., 2008, Springer.
  5. ASME, ASME Boiler and Pressure Vessel Code, Section III Rules for Construction of Nuclear Power Plant Components, Division 1, Subsection NB Class 1 Components, 2004, The American Society of Mechanical Engineers.
  6. Bannantine, J.A., Comer, J.J., Handrock, J.L., Fundamentals of Metal Fatigue Analysis, 1989, Prentice Hall, New Jersey.
  7. Heywood, R.B., Designing by Photoelasticity, 1952, Chapman & Hall Ltd., London.