ML042260221

Catawba Hearing 07/14/04 - Staff Exhibit (Official Exhibit #37), Professional Qualifications Statements of Undine Shoop, Ralph R. Landry, and Ralph O. Meyer - Rec'D 07/14/04

ML042260221
Person / Time
Site:	Catawba
Issue date:	07/14/2004
From:	Landry R, Meyer R, Undine Shoop Office of Nuclear Reactor Regulation
To:
Byrdsong A T
References
50-413-OLA, 50-414-OLA, ASLBP 03-815-03-OLA, Catawba-Staff-37, RAS 8299
Download: ML042260221 (11)

Text

Yes A sae 0-? 9l1 i Professional Qualifications 1 \\\

Undine Shoop DOCKETED Reactor Systems Engineer ISNRC US Nuclear Regulatory Commission August9,2004(11:45AM)

OFFICE OF SECRETARY \ c:

Education ADJUDICATIONS STAFF \\\

B.S., The Pennsylvania State University, Nuclear Engineering, 1994 l \ l M.S., The Pennsylvania State University, Nuclear Engineering, 1996 Experience Reviewed Lead Test Assembly (LTA) application for the use of down-blended weapons uranium fuel.

Completed multiple applications for insertion of LTAs Into current LWR cores. Some of these reviews involved requests to use the LTAs above current approved bumup limits.

Reviewed multiple exemption requests for the use of M5 cladding in current LWRs.

Led the Reactor Systems Branch team for the MOX LTA application review.

Completed multiple requests to change fuel parameters in the Technical Specifications of current LWRs.

Reviewed new fuel designs, both for generic and plant specific applications.

Primary author of a commission paper on the agency technical and regulatory needs prior to the use of MOX fuel commercial LWRs for the office of research.

Member of the ANS 19.6 Standards Working Group which has evaluated the PWR Startup Physics testing program.

While working at the Pennsylvania Power and Light Co.

Generated SIMULATE-E generic beginning of life rod withdrawal sequence input decks for reactivity evaluation of core startup following outages.

Developed and guided a beginning of life core asymmetry check program through the quality assurance program requirements.

Evaluated local power range monitor (LPRM) trends to identify outage replacement candidates. Performed transversing in-core probe surveillances.

Verified acceptability of new fuel during fuel receipt procedures.

etlcS= sacy-0 II,

DOCKETED USNRC August 9,2004 (1 1:45AM)

OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF

Publications uTRAC-BF1 THREE DIMENSIONAL BWR VESSEL THERMAL-HYDRAULIC and ANSYS Stress ANALYSIS FOR BWR CORE SHROUD CRACKING" in Annuals of Nuclear Energq, Vol. 25, No. 1-3, pp. 65-81, 1998.

Professional Affiliations Founding member of a new professional society, North American Young Generation in Nuclear, Incorporated December 1999. Currently semving as the Past Prosident.

Member of the American Nuclear Society from 1992 to the present.

Chairman of the Professional Woman in ANS Committee, member from 1994 to the present.

• Chaiiaii the Nuclear Inations Safety Division Progrm i

!he AS45

Professional Qualifications RALPH R. LANDRY PROFESSIONAL EXPERIENCE October 1995 - Present: Senior Reactor Engineer (Nuclear) in the Reactor Systems Branch, DSSA, Office of Nuclear Reactor Regulation, NRC. Duties include:

Group Leader for review of Siemens EXEM/PWR code, EPRI RETRAN-3D code, Framatome ANP S-RELAP5 code, GE update to SAFER/GESTR code, and the GE TRACG code. Lead reviewer and principle author of the Safety Evaluation Report of ESBWR TRACG application review. Review of the thermal hydraulic and analysis computer codes (LOFTRAN and NOTRUMP) for the Westinghouse AP600 advanced, passive reactor design. Work involves interaction with the applicant as well as management of contractor support and presentations before the ACRS. Assistance with the review of steam generator degradation problems experienced at the Combustion Engineering NSSS reactor sites. Inspection of the Yankee Atomic Electric Company LOCA analysis code pertaining to the Vermont Yankee Nuclear Plant. Inspection resulted in violations, enforcement action and civil penalties. Inspection of the Westinghouse LOCA Engineering Review Committee and Westinghouse analysis methods for the Westinghouse NSSS operating reactor plants. Inspection of the Siemens Power Corporation fuel and LOCA code methodologies. Member of Safety System Functional Inspection team for Crystal River Unit 3.

October 1994 - October 1995: Senior Reactor Engineer (Nuclear) in the Advanced Reactor Project Directorate, Office of Nuclear Reactor Regulation, NRC. Duties included:

Thermal hydraulic and analysis computer code review of the CANDU standard design submittal. Work Involves interaction with the applicant and the Atomic Energy Control Board of Canada. Development and management of technical assistance contracts in support of reactor physics and thermal hydraulic analysis code reviews. Assistance to the Office of Information Resource Management in design, specification, procurement and scheduling for the Technology Discovery Center. Mentor for NRC Fellow during his one year stay with PDAR.

January 1992 - October 1993: Senior Reactor Engineer (Nuclear) In the Reactor Systems Branch, Detailed to the Analytic Support Group, Division of Systems Safety & Analysis, Office of Nuclear Reactor Regulation, NRC. Duties included:

Acquisition of engineering workstations for the Analytic Support Group's use in performing analysis computer code calculations. Workstations were acquired and networking and access to the Internet for the workstations was implemented.

Administration of the engineering workstation network as backup to the IRM System Administrator. Installation and testing of computer codes to be used by the Analytic Support Group. Codes installed include: SCDAP/RELAP5/MOD3, the Nuclear Plant Analyzer, CONTAIN, MELCOR, CONTEMPT/LT28, COGAP, COMPARE, VIPRE, and MINET. Support of the Advanced Reactor Directorate in installation of the CATHENA code for analysis of the CANDU reactor. Analysis of the Shearon Harris Nuclear Power Plant degraded high pressure injection safety injection system condition. Analysis of the

steam generator tube rupture coincident with main steam line break generic issue.

Evaluation of the AP600 advanced passive reactor design. Management of training contract for computer systems and code use training of Analytic Support Group staff.

Development and management of contracts with national laboratories and universities to provide support to the Analytic Support Group.

April 1991 - January 1992: Senior Reactor Engineer (Nuclear) in the Reactor Systems Branch, Division of Systems Technology, Office of Nuclear Reactor Regulation, NRC. Duties included:

Application of advanced thermal-hydraulic analysis computer codes to evaluation of advanced reactor designs, including the AP600 and SBWR designs. Analyses are also performed in support of the design testing programs. Evaluate capability of analysis codes to adequately perform licensing safety reviews of advanced reactor designs.

Develop capability within the branch to perform analyses of advanced designs. Includes determination of code needs, computer hardware configurations and training necessary to operate the required computer equipment and analysis codes. Develop the training program for the staff that will be performing the in-house reactor design analyses.

November 1989 - April 1991: Senior Research Engineer in the Accident Evaluation Branch, Office of Nuclear Regulatory Research, NRC. Duties included:

Program manager for the OECD/NEA TMI Vessel Integrity Project. This was a cooperative international program including the U.S. and ten partner countries under the auspices of the OECD/NEA. Preparation of revision to Severe Accident Research Program plan to include long term research plans. Program manager for lower head failure analysis research at INEL.

April 1987 - November 1989: Reactor Engineer in the Advanced Reactors Branch, Office of Nuclear Regulatory Research, NRC. Duties includedd:

Program manager for the PRISM and SAFR liquid metal reactor conceptual design reviews. Principal author of NUREG-1 368, the PRISM Safety Evaluation Report.

Principal author of NUREG-1 369, the SAFR Safety Evaluation Report. Assist with development of Standardization Rule, 10 CFR 52. Preparation of Commission paper regarding proposed review of PIUS reactor design.

September 1986 - April 1987: Nuclear Engineer in the Regulatory Improvements Branch, Office of Nuclear Reactor Regulation, NRC. Duties includedd:

Preparation of Generic Letter for Individual Plant Examinations for implementation of the NRC's Severe Accident Policy Statement. Review of Guidelines and Criteria prepared for the NRC by a national lab for the Large Dry PWR containment design. Preparation of the implementation of the Severe Accident Policy Statement guidance for new and future plant designs. Preparation of the Severe Accident section for the NRC 1986 Annual Report. Preparation of letters and materials pertaining to international cooperative work for the Director and Deputy Director, Division of Safety Review and Oversight. Review and comment, including preparation of input, on an international report on approaches to nuclear safety for the Director, NRR.

September 1984-September 1986: Administrator, Nuclear Energy Agency, Organization for Economic Cooperation and Development, Paris, France. On leave from the NRC with responsibilities for:

Coordination of international cooperative research in nuclear reactor thermal hydraulics and fuel behavior. Definition of recommended procedures for thermal hydraulic analysis computer code assessment and validation, including criteria for successful completion of code assessment. Coordination of International Standard Problem exercises in thermal hydraulics, containment response, and fuel behavior. Coordination of the international cooperative research programs in the Loss-of-Fluid Test facility in the United States, and the Halden Reactor Project in Norway. Development of an international cooperative effort for examination of the debris material from the Three Mile Island Unit 2 facility, and standard problem analyses of the TMI-2 accident.

The work resulted in completion of an international code validation matrix for PWR and BWR analysis codes, completion of four standard problem exercises, definition of five future standard problem exercises, and publication of State-of-the-Art reports on BWR Pressure Suppression Containment Systems, and PWR Fuel Behavior Under LOCA Conditions. This work was based on coordination of the work in twenty OECD/NEA member countries.

June 1982-September 1984: Program Manager, Office of Nuclear Regulatory Research, NRC.

Duties included:

Management of the Semiscale Project at the Idaho National Engineering Laboratory, Idaho Falls, Idaho. Management of the RELAP5 code development project at the INEL.

Management of the PWR code assessment program at the INEL. Development of the MB-2 steam generator research program at the Westinghouse Tampa Facility, with Westinghouse and EPRI. Participation in the Test Advisory Group, comprised of NRC, B&W, B&W Owners Group, and EPRI studying the issues surrounding the B&W reactor design, leading to development of a research facility and program to resolve the identified needs for the B&W plant design. Preparation of the Integral Systems Tests sections of the NRC Annual Report and the Research Office's Long Range Research Plan.

May 1978-June 1982: LOFT Research Branch, Office of Nuclear Regulatory Research, NRC.

Duties included:

Program management and technical direction for heat transfer and fluid dynamics analysis as related to Integral systems experiments. Review of results of research and test programs to determine progress and to assure that the work is applicable to analysis methods. Assist the Assistant Director for Water Reactor Safety Research in developing research goals and objectives for the Division programs. Provide liaison and program management for the U.S. and international ECCS Standard Problem programs.

Budget management of the LOFT Program; $54M per year.

January 1976-May 1978: Reactor Engineer, Reactor Systems Branch, Division of Operating Reactors, Office of Nuclear Reactor Regulation, NRC. Duties included:

Review of safety analyses associated with operating reactor fuel reloads. Safety evaluation of ECCS redesign work at San Onofre Unit 1. Evaluation of plant modifications for BWRs to meet ATWS requirements. Evaluation of isolation capability of PWR low pressure systems from the high pressure reactor coolant system.

February 1974-January 1976: Reactor Engineer, Reactor Systems Branch, Division of Technical Review, Office of Nuclear Reactor Regulation, NRC. Duties included:

Review of license applications, particularly the Westinghouse Standard Plant designs RESAR-41, RESAR-3S, and the Westinghouse 17x1 7 core design. Review of license application for the South Texas Project. Development of Branch Position on isolation capability and requirements for PWR Residual Heat Removal Systems. ECCS evaluation model review for acceptance under 10 CFR 50.46.

February 1972-February 1974: Nuclear Engineer, Bechtel Power Corporation. Duties included:

Analysis of containment design and subcompartment pressurization response for the Calvert Cliffs, Davis Besse, and Farley plants. Evaluation of control of post-LCOA hydrogen generation in BWRs. Development of hydrogen generation and transport code for BWRs. Development of containment and containment subcompartment thermal hydraulic codes. Evaluation of radiological consequences of all design basis accidents.

EDUCATION BS in Mechanical Engineering, University of Missouri-Rolla.

PhD in Nuclear Engineering, University of Missouri-Rolla. Dissertation: A Study of the Effect of Rotation on the Nucleate Boiling from a Vertical Copper Cylinder.

Professional Qualifications Ralph 0. Meyer U.S. Nuclear Regulatory Commission Washington, DC 20555 (301) 415-6789 rom~nrc.gov AREA OF EXPERTISE: Reactor Fuel Behavior EDUCATION: B.S. Physics, University of Kentucky, 1960 Ph.D. Physics, University of North Carolina, 1966 PROFESSIONAL EXPERIENCE:

1965-1968 Research Associate, Department of Physics University of Arizona Tucson, AZ 85710

[Experimental studies of mechanisms of diffusion in metals]

1968-1973 Assistant Metallurgist, Materials Science Division Argonne National Laboratory Argonne, IL 60439

[Experimental studies of fission gas bubble migration and plutonium migration in MOX fuel]

1973-Present Senior Technical Advisor, Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 (Formerly U.S. Atomic Energy Commission, various positions held)

[Approx. 10 years in Reactor Fuels Section on licensing (section leader);

10 years in research on severe accidents and advanced reactors (all fuel related); 10 years in research on fuel behavior under conditions of design-basis accidents]

ACHIEVEMENTS: Phi Beta Kappa Sigma Xi NSF Cooperative Graduate Fellowship NRC Meritorious Service Award (Other NRC Awards)

More than 40 Published Papers and Reports Listed, American Men and Women of Science Listed, Who's Who in Technology Member, American Nuclear Society

PUBLICATIONS:

1. H. P. Layer and R. 0. Meyer, "Improved Precision Chuck for Sectioning Diffusion Samples," Rev. Sci. Instr. 33,1458 (1962).

2. R. 0. Meyer and L. Slifkin, "Composition and Vacancy Flux Effects in Tracer Diffusion in Silver-Gold Alloys," in Thermodynamics (International Atomic Energy Agency, Vienna, 1966), Vol. II, p. 297.

3. R. 0. Meyer and L. M. Slif kin, 'Activity Coefficient and Vacancy-Flow Effects on Diffusion in Silver-Gold Alloys," Phys. Rev. 149, 556 (1966).

4. R. 0. Meyer, "Activity-Coefficient and Vacancy Flow Effects on Tracer Diffusion Coefficients in Silver-Gold Alloys," Phys. Rev. 163, 641 (1967).

5. R. 0. Meyer and H. Sekizawa, "Capacitor Method for Measuring Lathe Section Thicknesses Applicable Primarily to Diffusion Studies," Rev. Scl. lnstr. 39, 265 (1968).

6. R. 0. Meyer, "Pressure and Vacancy-Flow Effects on the Kirkendall Shift in Silver-Gold Alloys," Phys. Rev. 181, 1086 (1969).

7. R. 0. Meyer and J. C. Voglewede, "Temperature Gradient Vacuum Furnace for Diffusion Studies to 20000C," Rev. Sci. Instr. 42, 993 (1971).

8. M. H. Greene, A. P. Batra, R. C. Lowell, R. 0. Meyer, and L. M. Slifkin, "Activity Coefficient and Vacancy Flux Effects on Tracer Diffusion in Silver-Gold Alloys," Phys.

Stat. Sol. 5, 365 (1971).

9. R. 0. Meyer and B. J. Buescher, "A Simple Method of Calculating the Radial Temperature Distribution in a Mixed-oxide Fuel Element," Nucl. Technol. 14,153 (1972).

10. R. 0. Meyer, E. M. Butler, and D. R. O'Boyle, "Actinide Redistribution in Mixed-oxide Fuels Irradiated in a Fast Flux," ANL-7929, May 1972.

11. D. R. O'Boyle and R. 0. Meyer, "Redistribution of Uranium and Plutonium in Mixed-Oxide Fuels During Irradiation," Behavior and Chemical State of Irradiated Ceramic Fuels (International Atomic Energy Agency, Vienna, 1974), p. 41.

12. E. M. Butler and R. 0. Meyer, "Diffusion of Plutonium and Uranium in Irradiated Mixed-oxide Fuel," J. Nucl. Mater. 47, 229 (1973).

13. R. 0. Meyer, D. R. O'Boyle, and E. M. Butler, "Effect of Oxygen-to-metal Ratio on Plutonium Redistribution in Irradiated Mixed-oxide Fuel," J. Nucl. Mater. 47, 265 (1973).

14. B. J. Buescher and R. 0. Meyer, "Thermal-gradient Migration of Helium Bubbles in Uranium Dioxide," J. Nucl. Mater. 48,143 (1973).

15. R. 0. Meyer, E. M. Butler, and D. R. O'Boyle, "Actinide Redistribution in Mixed-oxide Fuels Irradiated in a Fast Flux," ANL-7929 Supplement 1, July 1973.

16. R. 0. Meyer, "Analysis of Plutonium Segregation and Central Void Formation in Mixed-oxide Fuels," J. Nucl. Mater. 50,11 (1974).

17. R. 0. Meyer, "PLUTO2 -- A Plutonium Segregation and Restructuring Subroutine," ANL-8043, October 1973.

18. R. 0. Meyer, "Comments on Several Papers by Olander Regarding Actinide Redistribution in Mixed-oxide Fuels," J. Nucl. Mater. 52, 326 (1974).

19. (R. 0. Meyer and P. M. Wood),' "Effects of Plutonium Utilization on the Performance of Light Water Reactors," WASH-1 303, April 1974.

20. R. 0. Meyer, C. R. Hann, and D. D. Lanning, "Effects of Fissile Atom Segregation in Light Water Reactor Plutonium Recycle Fuels," Nucl. Technol. 27, 389 (1975).

21. (R. 0. Meyer),' "The Role of Fission Gas Release in Reactor Licensing,' NUREG-751077, November 1975.

22. R. 0. Meyer, "The Analysis of Fuel Densification,' NUREG-0085, July 1976.

23. R. 0. Meyer, C. E. Beyer, and J. C. Voglewede, "Fission Gas Release from Fuel at High Burnup," NUREG-0418, March 1978.

24. R. 0. Meyer, C. E. Beyer, and J. C. Voglewede, "Fission-Gas Release from Fuel at High Bumup," Nucl. Safety 19, 699 (1978).

25. Ralph 0. Meyer, 'Authors Response to the Preceding Letter," Nucl. Safety 20, 421 (1979).

26. D. A. Powers and R. 0. Meyer, "Evaluation of Simulated-LOCA Tests that Produced Large Fuel Cladding Ballooning," NUREG-0536, March 1979.

27. D. A. Powers and R. 0. Meyer, 'Cladding Swelling and Rupture Models for LOCA Analysis," NUREG-0630, April 1980.

28. S. B. Hosford, R. Mattu, R. 0. Meyer, E. D. Throm, and C. G. Tinkler, 'Asymmetric Blowdown Loads on PWR Primary Systems," NUREG-0609, January 1981.

29. R. 0. Meyer, L. D. Noble, C. S. Rim, C. E. Beyer, R. L. Ritzman, M. J. F. Notley, and R.

A. Lorenz, "Background and Derivation of ANS-5.4 Standard Fission Product Release Model," NUREG/CR-2507, January 1982.

30. M. R. Kuhlman, D. J. Lehmicke, and R. 0. Meyer, "CORSOR User's Manual,'

NUREG/CR-4173, March 1985.

31. M. Silberberg, J. A. Mitchell, R. 0. Meyer, and C. P. Ryder, 'Reassessment of the Technical Bases for Estimating Source Terms," NUREG-0956, July 1986.

'During a brief period, agency policy (later reversed) prohibited the identification of authors on agency reports.

32. D. E. Carlson and R. 0. Meyer, Assessment of Databases and Modeling Capabilities for the CANDU 3 Design," NUREG-1 502, July 1994.

33. R. 0. Meyer, R. K. McCardell, H. M. Chung, D. J. Diamond, and H. H. Scott, "A Regulatory Assessment of Test Data for Reactivity-initiated Accidents," Nuclear Safety, 37, 271 (1996).

34. R. Meyer, "Summary of High Bumup Fuel Issues -and NRC's Plan of Action,"

Proceedings of the Twenty-Fourth Water Reactor Safety Information Meeting (Washington 1996), NUREG/CP-0157,1997, pp. 79-82.

35. R. 0. Meyer, R. K. McCardell, H. M. Chung, D. J. Diamond, and H. H. Scott, UA Regulatory Assessment of Test Data for Reactivity-Initiated Accidents," Nuclear Safety, Vol. 37,1996, pp. 271-288.

36. R. 0. Meyer, R. K. McCardell, H. H. Scott, UA Regulatory Assessment of Test Data for Reactivity Accidents," Proceedings of the International Topical Meeting on Light Water Reactor Fuel Performance (Portland 1997), Amer. Nucl. Soc., 1997, pp. 729-744.

37. R. 0. Meyer, "NRC Activities Related to High Bumup, New Cladding Types, and Mixed-Oxide Fuel," Proceedings of an International Topical Meeting on Light Water Reactor Fuel Performance (Park City 2000), Amer. Nucl. Soc., 2000, pp. 736-744.

38. R. Meyer, "NRC Program for Addressing Effects of High Bumup and Cladding Alloy on LOCA Safety Criteria," Proceedings of the Topical Meeting on LOCA Fuel Safety Criteria (Aix-en-Provence, 2001), NEANCSNI/R(2001)18, 2001, pp. 65-74.

39. R. 0. Meyer, "Implications From the Phenomenon Identification and Ranking Tables (PIRTs) and Suggested Research Activities for High Bumup Fuel," NUREG-1749, September 2001.

40. G. M. O'Donnell, H. H. Scott, and R. 0. Meyer, UA New Comparative Analysis of LWR Fuel Designs," NUREG-1754, December 2001.

41. R. 0. Meyer, "LOCA Ductility Tests," Proceedings of the 2002 Nuclear Safety Research Conference (Washington), NUREG/CP-01 80, 2003, PP.99-108.

42. J. R. Strosnider, Jr., and R. 0. Meyer, "Safety Research in a Competitive World,"

Proceedings of ENS TopFuel2003 (WOrzburg 2003), European Nucl. Soc., 2003, Opening Session.

43. R. 0. Meyer, "A Scaling Method for RIA Data," Proceedings of the 2003 Nuclear Safety Research Conference (Washington), NUREG/CP to be published 2004.

The above are journal articles and formal reports. Additional conference presentations and seminars (e.g., recent seminars at MIT, Penn. State Univ., and N.C. State Univ.) have not been included.