Reactor Vessel Integrity - Fracture Toughness Criteria

News & Views, Volume 50 | Reactor Vessel Integrity


By:  Tim Griesbach and Dan Denis

Reactor Vessel Integrity - Fracture Toughness CriteriaThe 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.


News & View, Volume 46 | Baffle-Former Bolt Management- Cost:Benefit Studies

News & Views, Volume 46 | Baffle-Former Bolt Management: Cost/Benefit Studies

By:  Tim Griesbach and News & View, Volume 46 | Baffle-Former Bolt Management- Cost:Benefit Studies

For the past several years baffle-former bolt (BFB) cracking in pressurized water reactors has become a significant concern for of PWR plants. In 2016, three similar Westinghouse designed plants (Indian Point 2, Salem 1, and D. C. Cook Unit 2) experienced significant numbers of cracked BFBs, attributed to irradiation-assisted stress corrosion cracking (IASCC). These plants had common characteristics that included the 4-loop plant design, downflow configuration, and Type 347 stainless steel bolting material. BFB cracking is not an entirely new phenomenon as it was initially detected in the French PWR fleet in the 1990s. However, the extent of cracking found in some of the US plants has greatly exceeded prior cracking. Extensive industry programs have identified and categorized by tier group the most susceptible plants, and the EPRI Materials Research Program (MRP) has published guidance regarding baffle-former bolt UT inspections for PWR plants for detection of degraded and cracked bolts in the baffle-former assembly (MRP-2017-009).