Assessment of U.S. Embrittlement Trend Equations Considering the Latest Available Surveillance Data

2004 ◽  
Author(s):  
William L. Server ◽  
Randy G. Lott ◽  
Stan T. Rosinski
Keyword(s):  
Statistical Test ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Correlation Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Correlation Models ◽  
Pressurized Thermal Shock ◽  
Surveillance Programs

The mechanistically-guided embrittlement correlation model adopted in ASTM E 900-02 was based on a database of U.S. surveillance results current through calendar year 1998. There exists now an extensive amount of new surveillance data that includes a large amount of boiling water reactor (BWR) results from an integrated, supplemental surveillance program designed to augment the plant-specific BWR surveillance programs. These recent data allow a statistical test of the ASTM E 900-02 embrittlement correlation, as well as the NRC correlation model currently being used in the pressurized thermal shock (PTS) re-evaluation effort and the older Regulatory Guide 1.99, Revision 2 correlation. Even though the ASTM E 900-02 embrittlement correlation is a simplified version of the NRC model, a comparison of the two embrittlement correlation models utilizing the new database has proven to be revealing. Based on the new BWR data, both models are inadequate in their ability to predict BWR results; this inadequacy has even more significance for extrapolation outside of the database for BWR heat-up and cool-down curves, as well as some pressurized water reactor (PWR) heat-up curves. Other aspects of the two models, as revealed from this preliminary look at the new data, are presented.

U.S. High Fluence Power Reactor Surveillance Data—Past and Future

2014 ◽  
Vol 136 (2) ◽  
Author(s):  
William L. Server ◽  
Timothy C. Hardin ◽  
J. Brian Hall ◽  
Randy K. Nanstad
Keyword(s):  
Prediction Models ◽  
Power Reactor ◽  
Degradation Mechanism ◽  
Surveillance Data ◽  
Surveillance Program ◽  
High Fluence ◽  
Operating Life ◽  
Pressurized Water ◽  
Water Reactor ◽  
Surveillance Programs

Enhanced radiation embrittlement at high fluence, indicative of extended operating life beyond 60 years for current operating pressurized water reactor (PWR) vessels, has been identified as a potential limiting degradation mechanism. Currently, there are limited U.S. power reactor surveillance data available at fluences greater than 4 × 1019 n/cm2 (E > 1 MeV) for comparison with existing embrittlement prediction models. Additional data will be required to support extended operations to 80+ years, where some plants are projected to have peak vessel fluences approaching 1 × 1020 n/cm2. A number of programs are designed to contribute to the high fluence surveillance data to support extended operating life. The U.S programs include the Coordinated PWR Reactor Vessel Surveillance Program (CRVSP), the PWR Supplemental Surveillance Program (PSSP), and the Light Water Reactor Sustainability (LWRS) Program. The LWRS Program involves generation of high fluence test reactor data on many different reactor pressure vessel steels and model alloys, including some of the same PWR vessel materials irradiated to higher fluences in conventional power reactor surveillance programs. This paper surveys the existing high fluence data and the data projected to come from the above listed programs to show when such data will become available. The data will be used to validate or revise embrittlement trend correlations applicable for the high fluence regime. Mechanical property data are being developed, and fine-scale microstructure data are being produced using state-of-the-art methods.

Sign in / Sign up

Export Citation Format

Share Document