Mechanical property variation within Inconel 82/182 dissimilar metal weld between low alloy steel and 316 stainless steel

Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment

The response of alloy 690 tubing in a pressurized water reactor environment

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the Code specify fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of fatigue crack initiation in austenitic stainless steels in LWR coolant environments. The existing fatigue {var_epsilon}-N data have been evaluated to establish the effects of key material, loading, and environmental parameters (such as steel type, strain range, strain rate, temperature, dissolved-oxygen level in water, and flow rate) on the fatigue lives of these steels. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. The influence of reactor environments on the mechanism of fatigue crack initiation in these steels is also discussed.

/pdf/mechanism-and-estimation-of-fatigue-crack-initiation-in-1mfkp229hg.pdf

Mechanism and estimation of fatigue crack initiation in austenitic stainless steels in LWR environments.

In probabilistic safety assessments of pressurized water reactors, it is imperative to assess the potential and frequency of steam generator tube ruptures. Estimation of the frequency of steam generator tube ruptures has traditionally been based on historical occurrences, which are not applicable to new designs of steam generators with different geometries, material properties, degradation mechanisms, and thermal-hydraulic behaviors. This paper presents a new probabilistic mechanistic-based approach for estimating steam generator tube rupture frequency that is based on the principle that the failure of passive systems is governed by degradation or unfavorable conditions created through the underlying operating conditions and underlying mechanical, electrical, thermal, and chemical processes. This developed approach identifies, probabilistically models, and simulates potential degradations in new and existing steam generator designs to assess degradation versus time, until such degradation exceeds a known endurance limit. An example application of the proposed reliability prediction approach is presented for a new design of small modular reactor steam generators consisting of helically coiled tubes fabricated with advanced tube materials. This developed probabilistic physics-of-failure―based approach, when combined with probabilistic safety assessment techniques, can provide an effective tool for the evaluation of the safety and reliability of steam generators, particularly new steam generator designs used in advanced reactors.

A Probabilistic Physics-of-Failure Approach to Prediction of Steam Generator Tube Rupture Frequency

Reactor vessel internal components made of nickel-base alloys are susceptible to environmentally assisted cracking (EAC). A better understanding of the causes and mechanisms of this cracking may permit less conservative estimates of damage accumulation and requirements on inspection intervals. The objective of this work is to evaluate and compare the resistance of Alloys 600 and 690 and their welds, such as Alloys 82, 182, 52, and 152, to EAC in simulated light water reactor environments. The existing crack growth rate (CGR) data for these alloys under cyclic and constant loads have been evaluated to establish the effects of alloy chemistry, cold work, and water chemistry. The experimental fatigue CGRs are compared with CGRs that would be expected in air under the same mechanical loading conditions to obtain a qualitative understanding of the degree and range of conditions for significant environmental enhancement in growth rates. The existing stress corrosion cracking (SCC) data on Alloys 600 and 690 and Alloy 82, 182, and 52 welds have been compiled and analyzed to determine the influence of key parameters on growth rates in simulated PWR and BWR environments. The SCC data for these alloys have been evaluated with correlations developed by Scott and by Ford and Andresen.

/pdf/effects-of-alloy-chemistry-cold-work-and-water-chemistry-on-3st8ypjwmb.pdf

W. K. Soppet

Papers

Effects of alloy chemistry, cold work, and water chemistry on corrosion fatigue and stress corrosion cracking of nickel alloys and welds.

IASCC susceptibility of austenitic stainless steels and Alloy 690 in high dissolved oxygen water environment.