ML072780524

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Oyster Creek September 2007 Evidentiary Hearing - Applicant Exhibit 36, Curriculum Vitae of Dr. Har Mehta (Duplicative of Witness Curriculum Vitae That Forms Part of Exhibit D)
ML072780524
Person / Time
Site: Oyster Creek
Issue date: 09/20/2007
From:
GE-Hitachi Nuclear Energy Americas
To:
Office of Nuclear Reactor Regulation
SECY RAS
References
50-219-LR, AmerGen-Applicant-36, RAS 14247
Download: ML072780524 (7)
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Text

RA5 APPLICANT'S EXH. 36 Dr. Mehta Biography Dr. Mehta received his B.S. in Mechanical Engineering from Jodhpur University (India),

M.S. and Ph.D. from University of California, Berkeley. He was elected an ASME Fellow in 1999 and is a Registered Professional Engineer in the State of California.

Dr. Mehta has been with GE Nuclear Division (now called GE-Hitachi Nuclear Energy) since 1978 and currently holds the position of Chief Consulting Engineer, Mechanics.

He has over 30 years of experience in the areas of stress analysis, linear-elastic and elastic-plastic fracture mechanics, residual stress evaluation, and ASME Code related analyses pertaining to BWR components. He has also participated as principal investigator or project-manager for several BWRVIP, BWROG and EPRI sponsored programs at GE, including the Large Diameter Piping Crack Assessment, IHSI, Carbon Steel Environmental Fatigue Rules, RPV Upper Shelf Margin Assessment and Shroud Integrity Assessment. He is the author/coauthor of over 35 ASME Journal/Volume papers. Prior to joining GE, he was with Impell Corporation where he directed various piping and structural analyses.

For more than 20. years, Dr. Mehta has been an active member of the ASME Boiler &

pressure Vessel Code,Section XI Subgroup on Evaluation Standards and associated working and task groups. He also has been active for many years in ASME's PVP Division as a member of the Material & Fabrication Committee and as conference volume editor and session developer. His professional participation also included several committees of the PVRC, specially the Steering Committee on Cyclic Life and Environmental Effects in Nuclear Applications. He had a key role in the development of environmental fatigue initiation rules .that are currently under consideration for adoption by various ASME Code Groups.

U.S. NUCLEAR REGULATORY COMMISSION DOCKETED USNRC Doc, t No. Official ExhibitNo.

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DR. HARDAYAL S. MEHTA ACADEMIC QUALIFICATION B.E.,(Mechanical). 1964 University of Jodhpur (India)

M.S. (Mechanical) 1968 University of California, Berkeley Ph.D. (Mechanical) 1971 University of California, Berkeley LIST OF PUBLISHED TECHNICAL PAPERS AUTHORED/COAUTHORED BY H.S. MEHTA

1. E.R. Lambert, H.S. Mehta and S. Kobayashi, "A New Upper-Bound Method for Analysis of Some Steady-State Plastic Deformation Processes," Journal of Engineering for Industry, Trans. of ASME, Vol. 91, Series B, No.3, August 1969.

2.. H.S. Mehta, A.H. Shabaik and S. Kobayashi, "Analysis of Tube Extrusion," Journal of Basic Engineering, Trans. of ASME, Volume 92, Series B, No.2, 1970.

3. H.S. Mehta and S. Kobayashi, "Finite Element Analysis and Experimental Investigation of Sheet Metal Stretching," Journal of Engineering for Industry, Trans. of ASME, 1972.
4. H.S. Mehta and S. Ranganath, "Environmental Fatigue Crack Growth Analysis Based on Elastic-Plastic Fracture Mechanics," ASME Paper No. 82-PVP-23, 1982.
5. M.L. Herrera, H.S. Mehta and S. Ranganath, "Residual Stress Analysis of Piping with Pre-Existing Cracks Subjected to the Induction Heating Stress Improvement Treatment," ASME Paper No. 82-PVP-60, 1982.
6. S. Ranganath and H.S. Mehta, "Engineering Methods for the Assessment of Ductile FractureMargin in Nuclear Power Plant Piping," ASTM STP 803, Volume II, 1983, pp. 11-309-11-330.

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7. S. Ranganath, H.S. Mehta and D.M. Norris, "Structural Evaluation of Flaws in Power Plant Piping," PVP Vol. 94, Circumferential Cracks in Pressure Vessels and Piping-Vol. I, 1984, ASME.
8. Mehta, H.S., "An Assessment of the Probability of Double-Ended Guillotine Break in a BWR Recirculation Piping," Proceedings of International Conference on Nuclear Power Plant Aging, Availability Factor and Reliability Analysis; San Diego, CA, 1985, pp. 467-472.

Mehta, H.S., "J-Integral Analysis of Ductile Fracture Margin in Piping Weld Overlays,"

Transactions of the 9th SMiRT Conference; 1987, Volume G, pp. 469-474.

10. Mehta, H.S., "An Evaluation of Double-Ended Break and Leak Probabilities in a BWR Main Steam Piping System," Transactions of the 9th SMiRT Conference, 1987, Volume M, pp. 433-438.
11. H.S. Mehta, N.T. Patel and B. Chexal, "Application of Leak-Before-Break Justification Approach to BWR Piping," Proceedings of the American Power Conference, Volume.50, 1988, pp. 617-622.
12. R.L. Sindelar, N.G. Awadalla, N.P. Baumannand H.S. Mehta, "Life Extension Approach to the Reactor Vessel of a Nuclear Production Reactor," ASME PVP Volume 171, 1989, pp. 9-16.
13. H.S. Mehta, S. Ranganath, N.G. Awadalla, RL. Sindelar and G.R. Caskey, Jr., "Fracture Mechanics Evaluation of Savannah River Plant Reactors Considering Irradiated Toughness Properties," ASME PVP Volume 170, 1989, pp. 59-67.
14. N.G. Awadalla, R.L. Sindelar, W.L. Daugherty, Mehta, H.S. and S. Ranganath, "Leak-Before-Break Analysis of Type 304 Stainless Steel Piping,",Transactions of the 10th SMiRT Conference, 1989, Volume G, pp. 369-374.
15. W.L. Daugherty, N.G. Awadalla, R.L. Sindelar and H.S. Mehta, "A Failure Probability Estimate of Type 304;Stainless Steel Piping," Proceedings of ANS International Topical Meeting on The Safety, Status, and Future of Non-Commercial Reactors and Irradiation.

Facilities, 1990.

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16. H.S. Mehta and B. Chexal, "A Leak-Before-Break Assessment of BWR Recirculation Piping," PVP-Vol. 213/MPC-Vol. 32, ASME 1991, pp. 223-228.
17. H.S. Mehta, S.Ranganath, N.G. Awadalla, R.L. Sindelar, G.R. Caskey, Jr. and W.L.

Daugherty, "Development of J-Integral Based UT Flaw AcceptanceCriteria for Savannah River Reactor Tanks," Transactions of the 11 th SMiRT Conference, 19919, Volurne.G, pp.

19-24.

18. H.S. Mehta, W.L. Daugherty, N.G. Awadalla and R.L. Sindelar, "Double-Ended Break Probability Estimate for the 304 Stainless Steel*Main Circulation Piping of a Production
  • Reactor," Transactions of the 11th SMiRT Conference, 1991, Volume M, pp. 277-282.
19. H.S. Mehta, "An Assessment of the Frequency of Interfacing Loss of Coolant Accident due to Inadvertent Overpressurization of ECCS Piping," Transactions of the 11 th SMiRT Conference, 1991, Volume M, pp. 145-150.
20. H.S. Mehta and S. Ranganath, "An Environmental Fatigue Stress Rule for Carbon Steel Reactor Piping," ASME PVP Vol. 241 (Fatigue, Fractureand Risk),.pp. 17-23, 1992.
21. H.S. Mehta and S. Ranganath, "Management ofDegradation Mechanisms in Nuclear Power Plant Components," Contributed Paper for inclusion in the 'Decade 0f Progress in Pressure Vessel and Piping Technology' volume published in 1993.
22. H.S. Mehta, "Recent Progress in Structural Integrity Assessment Techniques for Components Subject to Service-Induced Degradation," Proceedings of the Second International Conference on Nuclear Engineering, San Francisco (1993).
23. H.S. Mehta, "A Low Upper Shelf Energy Fracture Mechanics Evaluation for a BWR Pressure Vessel," ASME PVP Vol. 260 (Fracture Mechanics: Applications and New Materials), pp. 59-64, 1993.
24. H.S. Mehta, "A Fracture Mechanics Evaluation of BWR Recirculation Pump Casing Considering Thermal Aging Embrittlement," ASME PVP Vol. 287 (Fracture Mechanics Applications), pp.93-104, 1994 3
25. H.S. Mehta and S.R. Gosselin, "An Enviornmental Factor Approach to Account for Reactor Water Effects in Light Water Reactor Pressure Vessel and Piping Fatigue Evaluations,"

ASME PVP Vol. .323, Fatigue and Fracture, pp. 171-185, 1996.

26. H.S. Mehta, "Application of EPRI/GE Environmental Factor Approach to Representative BWR Pressure Vessel and Piping Fatigue Evaluations," ASME PVP Vol. 360, Pressure Vessel and Piping Codes and Standards - 1998, pp. 413-425, 1998.
27. H.S. Mehta and S.R. Gosselin, "Enviornmental Factor Approach to Account for Water Effects in Pressure Vessel and Piping Fatigue Evaluations," Nuclear Engineering and Design, 181 (1998), pp. 175-197.
28. H.S. Mehta, "An Update on the EPRI/GE Environmental Fatigue Evaluation Methodology and Its Applications," ASME PVP Vol. 386, Probabilistic and Environmental Aspects of Fracture and Fatigue, pp. 183-193, 1999.
29. H.S. Mehta, "An Update on the consideration of Reactor Water Effects in Code Fatigue Initiation evaluations for Pressure Vessels andPiping," ASME PVP Vol. 410-2, Assessment Methodologies for Preventing Failure, pp. 45-51, 2000.
30. H.S. Mehta, "A Fracture Mechanics Evaluation of Service-Induced Flaws at Jet Pump Riser Elbow welds," ASME PVP Vol. 410-2, Assessment Methodologies for Preventing Failure, pp. 119-125, 2000.
31. H.S. Mehta, "A Fracture Mechanics Evaluation of Observed Cracking in a Recirculation Pump Shaft," ASME PVP Vol. 427,..Service Experience, Fabrication, Residual Stresses and Performance, pp. 3-13, 2001.
32. H.S. Mehta, R.M. Horn and G. Inch "A Fracture Mechanics Evaluation of Observed Cracking at a BWR-2 Reactor Pressure Vessel Weld;" ASME PVP Vol. 437, Service Experience and Failure Assessment Applications, pp. 153-164, 2002.
33. H.S. Mehta, G. Inch and Shashi Dhar"A Fracture Mechanics Evaluation of BWR Shroud Mid-Core Horizontal Weld to Justify Continued Operation," ASME PVP Vol. 463, Flaw Evaluation, Service Experience, and Reliability, pp. 178-190, 2003.

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34. H.S. Mehta, W.F. Miller and M.A. Brooks, "A Fracture Mechanics Evaluation of an Indication in an RPV Vertical Shell Weld Produced by the Upjohn Welding Technique,"

ASMIE PVP Volume 475; Flaw Evaluation, Service Experience, and Materials for Hydrogen Service; pp. 189-197, 2004.

35. A.R. Mehta, H.S. Mehta, H. Choe, G.B. Inch and R. Corieri, "Thermal-Hydraulic and LBB Evaluations to Justify Short Term Plant Operation With a CRD Return Line Susceptible to E

Potential Thermal Stratification," ASMIE PVP Volume 475; Flaw Evaluation, Service Experience, and Materials for Hydrogen Service; pp. 199-212, 2004.

36. H.S. Mehta and R.M. Horn, "An Assessment of Low Alloy Steel EAC Corrosion Fatigue Relationships for BWR Environment," Presentation at the 2005 ASMIE PVP Conference, July 17-2.1, 2005.
37. HS. Mehta "A Review of Fatigue & SCC Crack Growth rate Relationships for Ferritic &

Stainless Steels and Ni-Cr-Fe Materials in BWR Water Environment," Paper No. PVP2006-ICPVT11-93853, Proceedings of 2006 ASME PVP Conference.

38. H.S. Mehta and Henry H. Hwang, "Application of Draft Regulatory Guide DG-I 144 Guidelines for Environmental Fatigue Evaluation to a BWR Feedwater Piping System,"

Paper No. PVP2007-26143, Proceedings of 2007 ASME PVPConference.

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PARTICIPATION BY H.S. MEHTA IN ASME CODE, PVRC AND PVP DIVISION ACTIVITIES Member of the followingASME Section XI Code Groups:

Working Group on Pipe Flaw Evaluation Working Group on Flaw Evaluation Working Group on Operating Plant Criteria Subgroup on Evaluation Standards

2. Member of ASME Pressure Vessel & Piping (PVP) Division committees on Materials &

Fabrication and Codes & Standards.

3. Continued participation as Session Developer, Session Chairman at PVP Division Conferences. Edited three PVP conference volumes (Coeditor: PVP-Vol. 241* Fatigue, Fracture & Risk 1992; Principal Editor: PVP Vol. 260: Fracture Mechanics -Applications and New Materials, Principal Editor: PVP Volume 287: Fracture Mechanics Applications -

1994).

5. Member of PVRC Steering Committee on Cyclic Life and Environmental Effects in Nuclear Application-(200i-2004). This Steering Committee was considering the revision of ASMIE

.Code fatigue curves for low alloy, carbon and stainless steels to include environmental effects. Recommendations of this committee had significant impact on BWR Fatigue evaluations. As a part of this Committee, ! served as Chairman of Task Group on Total Damage Evaluation. I was also member of the several PVRC Working Groups/Task Groups which report to this Committee: W/G on S/N Analysis Data, T/G on Margins of Safety in Fatigue, T/G on Evaluation. Factors on Fatigue and W/G on da/dN Data Analysis..

6... Member, International Association of Structural Mechanics in Reactor Technology.

7. Member, ASTM (Committee E.08 - Fracture and Fatigue).

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