You are being redirected because this document is part of your ASTM Compass® subscription.
    This document is part of your ASTM Compass® subscription.


    Fatigue and Fracture Analysis of Type 316L Thin-Walled Piping for Heavy Water Reactors: Crack Growth Prediction Over 60 Years (With and Without Stratification) and Flawed Pipe Testing

    Published: Jan 1994

      Format Pages Price  
    PDF (740K) 23 $25   ADD TO CART
    Complete Source PDF (15M) 23 $131   ADD TO CART


    Thermal stresses in piping produce through-wall stress gradients that may influence crack growth in the pipe wall. In heavy water reactors (HWRs), the pressure-induced stress in piping is small due to the low pressure values. For major HWR plant transients, the thermal stresses (both wall profile and bending due to axial thermal growth) provide approximately 80% of the nominal pipe stress. Cases of local thermal stratification can provide a significant increase in the pipe stresses and the number of stress cycles applied to the piping. One focus in this paper will be to assess the impact of thermal stresses, in general, and thermal stratification, in particular, on the growth of flaws in a pipe due to fatigue loadings. The study will evaluate both the primary loop piping and the pressurizer surgeline piping. Evaluations of fatigue crack growth will be developed for both base material and weldments. The crack growth analysis for 60-year operation of a HWR plant will be reviewed. This analysis will be provided for the primary loop piping and the pressurizer surgeline piping and will establish the worst-case flaw size predicted from fatigue analysis by end of life.

    Full-scale flawed pipe fracture testing has been completed at the Oak Ridge National Laboratory (ORNL) in an approach to show the improbability of an instantaneous double-ended-guillotine break (DEGB) for HWR primary piping. This testing required a major facility (Pipe Impact Test Facility, PITF) to apply all possible design loads, including an equivalent major earthquake (called the SSE earthquake). The facility was designed and built at ORNL in six months. The test article, a 6.1-m (20-ft) long, 406-mm (16-in.) diameter, Schedule 40, 316 L Type stainless steel pipe was fabricated to exacting standards and inspections following the nuclear industry standard practices. A flaw was machined and fatigued into the pipe at a butt weld as an initial condition. The flaw/crack was sized to be beyond the worst-case flaw that HWR piping could see in 60 years of service—if all leak detection systems and if all crack inspection systems failed to notice the flaw's existence.

    Since October 1991, the test article was subjected to considerable overloadings. The pipe was impacted 104 times at levels equal to and well beyond the safe-shutdown earthquake (SSE) loadings. In addition, 569 751 fatigue cycles and 20 purposeful static overloads have been applied in order to extend the flaw to establish the data necessary to confirm fracture mechanics theories, and, more importantly, to simply demonstrate that an instantaneous DEGB is highly improbable for a HWR primary system at an operating pressure of 1.72 MPa and temperature of 110°C.


    fatigue crack growth, nuclear piping, cracks, stainless steels, fracture testing, fracture (materials), fatigue (materials), testing methods, test automation, data analysis

    Author Information:

    Poole, AB
    Senior Development Staff, Oak Ridge National Laboratory, Oak Ridge, TN

    Committee/Subcommittee: E08.06

    DOI: 10.1520/STP13964S

    CrossRef ASTM International is a member of CrossRef.