Environmental Science Division (EVS) a Division of Argonne National Laboratory

Effects of light water reactor coolant environment on the fatigue lives of reactor materials

July 8, 2013

A metal component can become progressively degraded, and its structural integrity can be adversely impacted when it is subjected to repeated fluctuating loads, or fatigue loading. Fatigue loadings on nuclear reactor pressure vessel components can occur because of changes in pressure and temperature caused by transients during operation, such as reactor startup or shutdown and turbine trip events.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code recognizes fatigue as a possible cause of failure of reactor materials and provides rules for designing nuclear power plant components to avoid fatigue failures. For various materials, the ASME Code defines the allowable number of fatigue cycles (or the fatigue life) for a given applied load. However, the ASME Code does not explicitly address the impact of light water reactor (LWR) coolant environments on fatigue life. The available fatigue data illustrate potentially significant adverse effects of coolant environments on the fatigue resistance of the steels used in the manufacture of pressure vessels and piping. For example, under certain environmental and loading conditions, fatigue lives can be shorter by as much as a factor of 15 in water versus air.

In a study sponsored by the U.S. Nuclear Regulatory Commission (NRC), EVS researchers have conducted a comprehensive review of the available fatigue test data for reactor pressure vessel and piping materials. The objective was to identify the material, environmental, and loading parameters that influence the initiation of fatigue cracking, as well as to establish the effects of key parameters on the fatigue life of the materials. Fatigue life decreases in water; the magnitude of the decrease depends on temperature, loading rate, reactor water chemistry, and – for some materials – sulfur content. In addition, the mechanism of fatigue crack initiation is different in air and LWR environments. A surface crack follows a zigzag pattern in air, whereas entirely straight cracking is observed in water (see images of component surface below). In air, fatigue cracks in carbon steels grow along relatively soft ferrite regions and avoid the hard pearlite regions, but in water the cracks appear to grow straight, perpendicular to the loading axis, through both the ferrite and pearlite regions (see images of cross section below). The straight cracks indicate a change in the mechanism of fatigue crack initiation, which significantly decreases the fatigue life of the material.

A surface crack follows a zigzag pattern in air, whereas entirely straight cracking is observed in water.
Fatigue crack pattern on the surface of the component. A surface crack follows a zigzag pattern in air (left), whereas entirely straight cracking is observed in water (right). [Source: Argonne National Laboratory]
In air, fatigue cracks in carbon steels grow along relatively soft ferrite regions and avoid the hard pearlite regions, but in water the cracks appear to grow straight, perpendicular to the loading axis, through both the ferrite and pearlite regions.
Fatigue crack pattern on a cross section of the material. In air (left), fatigue cracks in carbon steels grow along relatively soft ferrite regions and avoid the hard pearlite regions, but in water (right) the cracks appear to grow straight, perpendicular to the loading axis, through both the ferrite and pearlite regions. [Source: Argonne National Laboratory]

EVS has developed methodology for incorporating the effects of LWR environments into ASME Code Section III fatigue evaluations for the design of reactor components.

The results of this work are documented in several reports produced for the NRC. An updated report documenting the recent findings is currently under review with NRC staff. The EVS results serve as the technical basis for NRC Regulatory Guide (RG) 1.207, “Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components due to the Effects of the Light Water Reactor Environment for New Reactors.” RG 1.207 – and its anticipated update following approval of the NRC report currently under review – endorses the methodology developed by EVS for use in evaluating component fatigue lives for new reactors. The methodology is also applicable to fatigue analyses associated with license renewal applications for operating reactors. In addition, on the basis of the EVS studies and methodology, ASME Boiler and Pressure Vessel Code Case N-792, “Fatigue Evaluations Including Environmental Effects Section III, Division 1,” was prepared for performing fatigue evaluations of LWR coolant system and primary pressure boundary components when the effects of reactor coolant environment on fatigue life are judged to be significant.

Fatigue crack pattern along a vertical cross section of the material in a light water reactor environment
Fatigue crack pattern along a vertical cross section of the material in a light water reactor environment [Source: Argonne National Laboratory]
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