The ASME Boiler and Pressure Vessel (B&PV) Code Section XI Appendix C provides analytical procedures, criteria, and evaluation methodologies used to determine acceptability for continued service for a specified evaluation time period of flawed pipe. However, Appendix C applicability to subsurface flaws and flaws located on external pipe surfaces is unclear. Appendix C as currently written suggests surface flaws are (only) on the inner pipe diameter. It is recognized that flaw solutions specific to different combinations of the type of flaw, location on component, and failure mode may not be currently available. There are also inconsistencies in the equations for determining fracture toughness for ferritic piping between circumferential and axial-oriented flaws, and the allowable applied hoop stress definitions. Furthermore, there is recent work on several topics in Appendix C that necessitate updating Appendix C. Topics include stress intensity factor (SIF) solutions for circumferential and axial through-wall flaws in cylinders, and the method of combination of bending moments and torsion for elastic-plastic fracture mode and limit load analyses when the torsion stress does not exceed 0.2 times the flow stress. This paper summarizes the proposed ASME Code Section XI Appendix C revisions that will be incorporated in the 2017 edition of the Code. The impact of revising stress intensity factor solutions for circumferential and axial through-wall cracks in cylinders is also presented. In addition to technical changes, several errata are also suggested to be corrected.
As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.
When the Appendix G methodology, fracture toughness criteria for protection against failure, was first adopted by ASME Section III in 1972, it included a lower-bound Kir curve for ferritic steels with specified minimum room-temperature yield strength up to 50 ksi. In 1977, Section III Appendix G added a requirement to obtain fracture-toughness data for at least three heats (base metal, weld metal, and heat-affected zone) if the KIR curve is used for ferritic steels with specified minimum room-temperature yield strength between 50 and 90 ksi. The three-heat data requirement has not changed when the lower bound curve was adopted by Section XI, or when the lower-bound crack initiation toughness curve was changed from the dynamic Kir curve to the static KIc curve during the 2000s. Based on the accumulation of fracture-mechanics data of ferritic steels with specified minimum yield strength between 50 ksi and 90 ksi and their use for Class 1 pressure vessel production, Section XI recently expanded the applicability of the KIc curve to SA-508 Grade 2 Class 2, SA-508 Grade 3 Class 2, SA-533 Type A Class 2, and SA-533 Type B Class 2 whose specified minimum room-temperature yield strength is 65 ksi or 70 ksi. This paper describes the technical basis including the fracture-mechanics data to support the expansion of the applicability of the KIc curve by ASME Section XI.
Flaws detected in nuclear power plant components during in-service inspections are typically evaluated based on stress intensity factor influence coefficient databases and solutions from industry standards and public literature (e.g. API-579, ASME Section XI Code, WRC-175 Bulletin, Raju-Newman, etc). For certain components in the Pressurized Water Reactor (PWR) nuclear power plants, such as Bottom Mounted Instrumentation (BMI) nozzles, the cylindrical component geometry may fall outside the applicability limits of stress intensity factor influence coefficient databases. This situation occurs where the thickness to inner radius ratios of the cylindrical geometry is greater than 1.0. Accurate stress intensity factor (SIF) solutions are essential to flaw evaluation since the SIFs are used in the determination of both the allowable flaw size and crack growth in order to determine acceptability of the detected flaw. In this paper, stress intensity factor influence coefficients are generated based on a three-dimensional finite element analysis for axial flaws located on the inside surface and outside surface of a cylindrical component with thickness to inner radius ratios (t/Ri) of 1, 2, 4, & 6. Non-dimensional influence coefficients are determined at the deepest point of the crack front and the surface point of the flaw based on a 4th order polynomial fit for a through-wall stress profile. The influence coefficients are generated for semi-elliptical flaws with a/c ratios = 0.125 through 2; where a is depth of the elliptical flaw, and c is the half-length of the elliptical flaw. The influence coefficients developed are suitable for calculating stress intensity factors for cylindrical components with high thickness to inner radius ratios.
Reference fatigue crack growth (da/dN) curves for pressurized water reactor (PWR) environments have been proposed for ASME Section XI flaw evaluation applications in Code Case N-809. The reference curves are dependent on temperature, loading rate, mean stress, and cyclic stress range which are all contained in the da/dN model. This paper presents the application of N-809 in a fatigue crack growth analysis for a large diameter austenitic pipe in a PWR Reactor Coolant System main loop using the current analytical evaluation procedures in Appendix C of ASME Section XI. The example problem was used to evaluate the reference fatigue crack growth curves during the development of the code case and the results have been compared with other industry codes.
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