Session: DA-08-03 Practical Applications of FFS

Paper Number: 106451

106451 - Plant Validation of Flaw Evaluation Procedure for Axial Pwscc in Alloy 600 Reactor Vessel Head Penetration Nozzles 

The analytical flaw evaluation procedure of Subsubarticle IWB-3660 and Nonmandatory Appendix O of the ASME Boiler and Pressure Vessel Code, Section XI is available to PWR licensees to disposition the detection of primary water stress corrosion cracking (PWSCC) in Alloy 600 reactor vessel head penetration nozzles.  This procedure can be applied to accept a detected axial flaw for continued service if the flaw dimensions projected to the end of the evaluation period do not exceed the acceptance criteria of Table IWB-3663-1.  The crack growth prediction for future operation must consider the applicable loadings, including the tensile weld residual stresses that can dominate the stress state, as well as a PWSCC crack growth rate equation for Alloy 600 material.  This equation allows prediction of the PWSCC crack growth rate as function of metal temperature and crack-tip stress intensity factor.

In 2018, a U.S. PWR detected leakage emanating from an Alloy 600 reactor vessel closure head control rod/element drive mechanism (CRDM) penetration.  Retrospective review of the ultrasonic testing data collected over time revealed several operating cycles of size information for the nozzle axial PWSCC flaw that was determined to be the cause of the leakage.  The flaw initiated at the nozzle ID surface and grew from a measurable size to extend to the nozzle OD annulus over a period of about 10 years.  This experience is a unique opportunity to compare the observed detailed flaw size evolution with the PWSCC growth predicted for the earliest observed flaw dimensions per the Section XI flaw evaluation procedure as a validation exercise.

This paper documents the methods and results of this comparison.  Results are obtained applying the EPRI MRP-55 PWSCC crack growth rate equation first incorporated within ASME Section XI in 2010, as well as the more recent EPRI MRP-420 R1 equation.  These results show that the predicted time for crack growth extending through the nozzle wall is bounded by the greater actually observed time.  In addition, the PWSCC crack growth calculations are extended to assess the full implications of the observed evolution in flaw dimensions, including with regards to the detailed predicted stress field and the capability of the calculations to predict crack size and shape evolution.  As would be expected, the observed flaw growth tended to be greatest toward areas of greatest predicted tensile weld residual stress.  Furthermore, this additional investigation reveals that the usual simplifying approach applying the stress intensity factor solution for a pipe geometry results in a conservatively large estimate of the acceleration in flaw depth as the flaw approaches penetration through the nozzle thickness.  Improved agreement with the observed evolution in crack depth is obtained if the constraining effect of the J‑groove weld is approximated by increasing the pipe wall thickness.

Presenting Author: Kevin Fuhr Dominion Engineering, Inc.

Presenting Author Biography: During his time at DEI, Mr. Fuhr has focused primarily on materials aging and degradation issues in the nuclear power industry, including probabilistic and deterministic analysis of component aging. Since 2013, Mr. Fuhr has been involved with aging management issues for reactor primary loop components at operating nuclear power plants and for welded spent fuel canisters. Mr. Fuhr holds a B.S. and M.Eng. in Mechanical Engineering, both from Cornell University. He is a licensed professional engineer in the Commonwealth of Virginia.

Authors:

Kevin Fuhr Dominion Engineering, Inc.
Gideon Schmidt Dominion Engineering, Inc.
Glenn White Dominion Engineering, Inc.
Craig Harrington Electric Power Research Institute