Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Pressure-Temperature Limit Curve of Reactor Vessel by ASME Code Section III and Section XI

  • M.J. Jhung (Korea Institute of Nuclear Safety) ;
  • Kim, S.H. (Korea Institute of Nuclear Safety) ;
  • Lee, T.J. (Korea Institute of Nuclear Safety)
  • Published : 2001.10.01
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Abstract

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

Keywords

References

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