It is recognized that piping systems used in nuclear power plants have a significant amount of the safety margin, up to the point of boundary failure, even when the input seismic load exceeds the allowable design level. The reason is attributed to the large strength capacity of the piping systems in the plastic region. In order to establish an evaluation procedure, in which the inelastic behavior of piping systems is considered in a rational way, a task group activity under the Japan Society of Mechanical Engineers (JSME) has been conducted. As a deliverable of this activity, a Code Case in the framework of the JSME Nuclear Codes and Standards is now being developed. The Code Case provides the strain-based criteria, an evaluation procedure using the response-spectrum based inelastic analysis, and detailed inelastic response analysis based on a finite element model.
For developing the Code Case, inelastic benchmark and parametric analyses of the tests of a pipe element and piping system made of carbon steel were conducted to investigate the variation of the elastic-plastic analyses results. Based on these analytical results, it is assumed that setting the yield stress has a significant influence on the inelastic analytical results, while the work hardening modulus in the bi-linear approximation of the stress-strain curve has little influence. From the results of the parametric analyses, it is confirmed that the variation in the analytical results among the analysts would be reduced by having a unifying analysis procedure. In this paper, the results of the parametric analyses and the variation in the elastic-plastic analysis are discussed.
Round robin analysis to investigate sensitivity of analysis results to finite element elastic-plastic analysis variables for nuclear safety class 1 components under severe seismic load
