News & Views, Volume 50 | Reactor Vessel Integrity
FRACTURE TOUGHNESS CRITERIA
By: Tim Griesbach and Dan Denis
The integrity of the nuclear reactor pressure vessel is critical to plant safety. A failure of the vessel is beyond the design basis. Therefore, the design requirements for vessels have significant margins to prevent brittle or ductile failure under all anticipated operating conditions. The early vessels in the U.S. were designed to meet Section VIII of the ASME Boiler and Pressure Vessel Code and later Section III. ASME Section III included requirements for more detailed design stress analyses also included a fracture mechanics approach to establish operating pressure-temperature heatup and cooldown curves and to assure adequate margins of safety against brittle or ductile failure incorporating the nil-ductility reference temperature index, RTNDT. This index is correlated to the material reference fracture toughness, KIC or KIa.
Radiation embrittlement is a known degradation mechanism in ferritic steels, and the beltline region of reactor pressure vessels is particularly susceptible to irradiation damage. To predict the level of embrittlement in a reactor pressure vessel, trend curve prediction methods are used for projecting the shift in RTNDT as a function of material chemistry and fluence at the vessel wall. Revision 2 of this Regulatory Guide is being used by all plants for predicting RTNDT shift in determining heatup and cooldown limits and hydrostatic test limits.
There have been several industry initiatives to support optimization of examination requirements for various items/components (both Class 1 and Class 2 components) in lieu of the requirements in the ASME Code, Section XI.
As part of the general design criteria for nuclear power plants, the primary structures and systems of the plant must be designed to handle postulated accident events, including the dynamic effects of postulated pipe ruptures. For a Boiling Water Reactor, analyzed events include various accident conditions in the recirculation piping, including a Loss of Coolant Accident (LOCA). One postulated LOCA event is assumed to be an instantaneous double-ended guillotine break of the recirculation line. This event causes several loads to be imparted on the reactor vessel, attached piping, and reactor internal components. [Some loads such as jet impingement, annulus pressurization, and pipe whip impart loads on the outside of the reactor vessel and the attached piping.][ Other loads, including flow-induced drag and acoustic loads, transmit loads inside the vessel on critical components such as jet pumps, core shroud, and the shroud support structure.] Figure 1 shows the pipe and resulting loads.
