ML15222A821

ENT000617 - Curriculum Vitae of Timothy J. Griesbach

ML15222A821
Person / Time
Site:	Indian Point
Issue date:	08/10/2015
From:	Entergy Nuclear Operations
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28133, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15222A821 (4)

Text

7/2015 Timothy J. Griesbach Senior Associate Education M.S. Metallurgy and Materials Science, Case Western Reserve University (1974) B.S. Metallurgy and Materials Science, Case Western Reserve University (1972) Professional Associations American Nuclear Society, 1982 to present American Society of Mechanical Engineers, 1982 to present Professional Experience 2006 - Present Sr. Associate, Structural Integrity Associates, Inc. Mr. Griesbach is assisting utilities on reactor vessel integrity issues, materials degradation management programs, PWR reactor vessel internals and plant operating issues. 1993 - 2005 Director of Technical Services, ATI Consulting. Mr. Griesbach is internationally known for his expertise in the areas of reactor vessel embrittlement and vessel integrity management. Activities at ATI included component life assessment, life attainment and aging management strategies, ASME Code and regulatory issues concerning reactor vessel integrity, application of advanced fracture mechanics methods, utility cost control, and technical/management decision-making for long-term component life. 1982 - 1993 Manager, Component Reliability Program, Electric Power Research Institute (1985 - 1993). Directed major research initiative to develop remedial measures for managing reactor vessel embrittlement. Managed program to develop the on-line fatigue usage monitoring system for critical reactor system components. Project Manager, Nuclear Safety Analysis Center (1982 to 1985). Developed methodologies to resolve generic safety issues including pressurized thermal shock, BWR pipe cracking, steam generator tube cracking, and containment integrity. 1977 - 1982 Principal Engineer, Nuclear Plant Engineering, Combustion Engineering, Inc. Evaluated Response to NSSS components to severe thermal, pressure and dynamic loads. 1974 - 1977 Materials Engineer, Materials Engineering and Research Laboratory, Pratt and Whitney Aircraft. Developed diffusion bonding process for fabrication of jet turbine blades.

T. J. Griesbach Page 2 7/2015 Related Experience Chairman, ASME Section XI Working Group on Operating Plant Criteria, 1984 to present. Loaned employee to Commonwealth Edison, PWR Plant Engineering group, 1988. Participant, US/USSR Cooperative Program on Vessel Thermal Annealing, 1990 to 1993. Consultant to IAEA on Vessel Embrittlement and Materials Degradation, 1990 to 1996. Key Contributor to EPRI MRP Reactor Internals Issues Task Group, 2000 2015. Summary Mr. Griesbach has over 35 years of experience in materials behavior and structural integrity of major nuclear components. He specializes in technical consulting utilizing state-of-the-art technologies for mitigating and resolving material degradation concerns in nuclear reactor vessels, internals, piping, and other major components. Key accomplishments include: In charge of numerous EPRI projects to develop tools for managing aging effects such as fatigue and reactor vessel embrittlement in nuclear pressure vessels. Authored many reports and technical papers on maintaining structural integrity of nuclear plant components, materials toughness and embrittlement prediction models, databases for monitoring vessel material properties, and strategic planning for license renewal. Interfaced with utilities, NEI, and NRC to resolve critical industry issues such as Pressurized Thermal Shock in PWRs and Integrated Surveillance Programs for BWRs Project team leader for successful competitive bids on major industry projects for reactor vessel thermal annealing, PWR internals aging management for license renewal, BWRVIP Integrated Surveillance Program, and probabilistic methods for developing operating P-T limit curves. Conducted workshops and utility training classes for managing reactor vessel integrity, nuclear plant surveillance programs, fracture mechanics for nuclear applications, and calculating plant operating P-T limits. Participated in strategic planning sessions, and developed technical justifications for resolving key plant issues at Calvert Cliffs 1&2, H. B. Robinson 2, Kewaunee, Beaver Valley 1, D. C. Cook 1&2, McGuire 1&2, Catawba 1&2, and Farley Unit 2. Part of the team that developed industry strategic plan for aging management of PWR vessel internals under the EPRI Materials Reliability Program. Member of the ASME Section XI Standards Committee. Selected Publications 1. Simple Mixing Model for Pressurized Thermal Shock Applications Nuclear Engineering and Design, Volume 74, Issue 2, February 1983. pp 193197.

T. J. Griesbach Page 3 7/2015 2. Griesbach, T. J., RiccardellApplication of Fatigue Monitoring to the Evaluation of Pressurizer Surge LinesIssue 2, August 1991. pp 163176. 3. Dynamic Elastic-Plastic Behavior of Circumferential Cracks in a Pipe Subject to Seismic Loading ConditionsJournal of Pressure Vessel Technology, Vol. 105, No. 1, February 1983. pp 63-72. 4. Timothy J. Griesbach, Dilip Dedhia, David O. Harris, Nathaniel G. Cofie, Kyle Amberge and Aparna AlleshwaramThe Influence of Flow Strength and Fracture Toughness on the Computed Reliability of Thermally Aged Grade CF-8M Cast Austenitic Stainless Steel Piping, ASME 2014 Pressure Vessels and Piping Conference, Paper No. PVP2014-28089, Anaheim, CA (2014). 5. H. S. Mehta, G. L. Stevens; D. V. Sommerville; M. Benson; M. Kirk, T. J. Griesbach, and J. Treatment of Stresses Exceeding Material Yield Strength in ASME Code Section XI Appendix G Fracture Toughness Evaluations, ASME 2014 Pressure Vessels and Piping Conference, Paper No. PVP2014-28397, Anaheim, CA (2014). 6. The Effect of Irradiation and Aging on Pressure Vessel Steel EmbrittlementNuclear Plant Aging, Availability Factor, and Reliability Analysis, San Diego, CA, July 7, 1985, American Society for Metals, (1986). p. 375 7. Materials Aging Management Programs at Nuclear Power Plants in the United States, IAEA 2nd International Symposium on Nuclear Power Plant Life Management, October, 2007, Shanghai, China. 8. A Probabilistic Approach to Baffle Bolt IASCC Predictionsrd International Conference on Nuclear Power Plant Life Management (PLiM) for Long Term Operations (LTO), May, 2012, Salt Lake City, Utah, USA. 9. Fatigue Lifetime Monitoring in Power Plantsn Fatigue Lifetime Predictive 473. International Journal of Fatigue, 12 (1), 65 (1993). 10. Kuo, A.-Y., Riccardella, P. C., and Griesbach, T. J., Development and Usage of P-T Calculator a PC-Based Computer Program for Constructing P-T Limit Curvesof a Pressure Vessel and Piping Conference, Vol. 195: ASME, New York (1990). pp 141-148. 11. Aging Management Strategies for Pressurized Water Reactor Vessel Internalsand Piping Conference, Paper No. PVP2004-3055, July 2004. pp. 37 41.

T. J. Griesbach Page 4 7/2015 12. Opening and Extension of Circumferential Cracks in a Pipe Subject to Dynamic LoadsPaper F5/1, Structural Mechanics in Reactor Technology, SMiRT-5, Berlin, Germany, (1979). 13. Radiation Embrittlement Mechanistic Modeling 170. 14. Griesbach, T. J., and Server, W. L., A Consideration of Scatter in Radiation Damage Trend Effects of Radiation on Materials: 18th International Symposium, ASTM STP 1325, R. K. Nanstad, M. L. Hamilton, F. A. Garner, and A. S. Kumar, Eds., American Society for Testing and Materials, (1999). pp 500 507. 15. An Overview of Radiation Embrittlement Modeling for Reactor Vessel Steelst of Nuclear Reactor Pressure Vessel Steels: An International Review (Fourth Volume), ASTM STP 1170, L. E. Steele, Ed., American Society for Testing and Materials, (1993). pp 99 117.