Published March 1990 | Version v1
Report

Use of J/JM-R curves in assessing the fracture behavior of low upper shelf toughness materials

  • 1. David Taylor Research Center, Annapolis, MD (USA)
  • 2. Naval Academy, Annapolis, MD (USA)

Description

The objective of this investigation was to evaluate the use of small specimen J/JM-R curves in assessing the fracture resistance behavior of reactor vessels containing low upper shelf toughness weldments. As required by 10 CFR, Part 50, reactor vessel beltline materials must maintain an upper shelf Charpy V-Notch (CVN) energy of at least 50 ft-lbs throughout vessel life. If CVN values from surveillance specimens fall below this value, the utility must demonstrate to the NRC that the lower values will provide margins of safety against fracture equivalent to those required by Appendix G of the ASME Boiler and Pressure Vessel Code. This paper presents recommendations regarding the material fracture resistance aspects of this problem and outline an analysis procedure for demonstrating adequate fracture safety based on CVN values. It is recommended that the deformation formulation of the J-integral be used in the analysis described above. For cases where J-integral fracture toughness testing will be required, the ASTM E1152-87 procedure should be followed, however data should be taken to 50% to 60% of the specimen remaining ligament. Extension of the crack growth validity limits for J-R curve testing, as described in E1152-87, can be justified on the basis of a singularity zone analysis which shows an engineering basis for J-control to 25% to 40% of the specimen remaining ligament. If J-R curve extrapolations are required for the analysis, a simple power law fit to data in the extended validity region should be used. The example analysis performed for low upper shelf LINDE 80 weld material, showed required CVN values for a reactor vessel with a 7.8 inch thick wall ranging from 32 ft-lbs to 48 ft-lbs, depending on the magnitude of the thermal stress component

Additional details

Publishing Information

Imprint Title
Seventeenth water reactor safety information meeting: Proceedings. Volume 3, Primary systems integrity; plant performance, testing and analysis; piping and NDE [nondestructive examination]; plant aging
Imprint Pagination
529 p.
Journal Page Range
p. 53-68.
Report number
NUREG/CP--0105-Vol.3

Conference

Title
17. water reactor safety information meeting.
Dates
23-25 Oct 1989.
Place
Rockville, MD (USA).

Optional Information

Secondary number(s)
CONF-8910222--Vol.3.