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Research Papers: Operations, Applications & Components

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

[+] Author and Article Information
William L. Server

ATI Consulting,
6 Laurel Branch Drive,
Black Mountain, NC 28711

Timothy C. Hardin

Electric Power Research Institute,
3420 Hillview Avenue,
Palo Alto, CA 94304

J. Brian Hall

Westinghouse Electric Co., LLC,
1332 Beulah Road,
Pittsburgh, PA 15235

Randy K. Nanstad

Oak Ridge National Laboratory,
P.O. Box 2008,
Oak Ridge, TN 37831

Contributed by the Pressure Vessel and Piping Division of ASME for publication in the JOURNAL OF PRESSURE VESSEL TECHNOLOGY. Manuscript received September 8, 2013; final manuscript received November 29, 2013; published online January 30, 2014. Assoc. Editor: Kunio Hasegawa.

J. Pressure Vessel Technol 136(2), 021603 (Jan 30, 2014) (5 pages) Paper No: PVT-13-1154; doi: 10.1115/1.4026149 History: Received September 08, 2013; Revised November 29, 2013

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.

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References

U.S. Regulatory Commission, 2011, Title 10, Section 50.61, of the Code of Federal Regulations, “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” U.S. Government Printing Office, Washington, DC.
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Eason, E. D., Odette, G. R., Nanstad, R. K., and Yamamoto, T., 2007, “A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels,” ORNL/TM-2006/530, Oak Ridge National Laboratory.
U.S. Regulatory Commission, 2010, Title 10, Section 50.61a, of the Code of Federal Regulations, “Alternative Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” US Government Printing Office, Washington, DC.
Kirk, M. T., 2010, “A Wide-Range Trend Embrittlement Trend Curve for Western RPV Steels,” Proceedings of Fontevraud 7, Paper No. A-106/T1.
Server, W., Hall, B., Rosier, B., and Hardin, T., 2013, “Comparison of Radiation Embrittlement Prediction Models to High Fluence U.S. Power Reactor Surveillance Data,” Proceedings of the 2013 ASME Pressure Vessels & Piping Conference, PVP-87309.
Hall, B., Server, W., Hardin, T., and Burgos, B., 2013, “PWR Supplemental Surveillance Program,” Effects of Radiation on Nuclear Materials: 26th International Symposium, ASTM STP 1572, M.Kirk, and E.Lucos, eds., ASTM, West Conshohocken, PA.
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Figures

Grahic Jump Location
Fig. 1

Difference between EONY predictions and measured TTS as a function of fluence for combined test reactor and power reactor data, showing ±2σ bounds for the EONY predictions [6]

Grahic Jump Location
Fig. 2

PWR surveillance capsules >3 × 1019 n/cm2 that will be tested in the U.S. through 2025 [9]

Grahic Jump Location
Fig. 3

Surveillance data currently available and that which will be generated from the CRVSP and the PSSP as compared to the projected U.S. PWR RPV fluences

Grahic Jump Location
Fig. 4

Schematic depiction of the flux/fluence range for the ATR-2 experiment, showing overlap of existing test reactor data from the Irradiation Variables (IVAR) program, the BR2 reactor in Belgium, and the Japanese Material Test Reactor (JMTR) (courtesy of Professor G. R. Odette)

Grahic Jump Location
Fig. 5

APT maps of (a) nickel and manganese distributions and a blowup of a Mn–Ni precipitate in a copper-free 1.6 wt. % Ni-1.6 wt. % Mn model alloy irradiated to 1.8 × 1019 n/cm2 at high flux and 290 °C [15], and (b) sequences of 1-nm-thick atom map slices through a Mn–Ni–Si precipitate in the Ringhals Unit 4 low copper, high fluence weld metal [20]

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