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Mark A. Rinckel
EDUCATION B.S., Mechanical Engineering, Ohio State University 1981
M.E., Mechanical Engineering, University of Virginia 1988
EXPERIENCE
2013-Present Responsible Technical Manager/Supervisor (First Line Leader)
At present, Supervisor of Codes and Methods responsible for all analytical
codes used by the Nuclear Analyses Group. Applicable codes include, but
are not limited to, the following: LOCA/NON-LOCA-- RELAP5/MOD 2 B&W,
S-RELAP5, Gothic; Radiological—GALE, MNCNP, SCALE, Microshield,
ELISA, RADTRAN; T-H—VAGEN, Porthos; PRA—MAAP, Melcor;. Fluence—
DORT; Environmental—FLO-2D, SLOSH.
Joined the Nuclear Analyses Group in 2013 as Responsible Technical
Manager in charge of the radiological, environmental and thermal-hydraulic
groups. Responsibilities included management of groups supporting the EPR
(radiological Chapters 11, 12, and 15 and T-H Chapters 6 and 9) and support
for operating plants (B&W, WEC, GE and CE-designed plants.
Environmental support included flooding hazards evaluations as required by
NRC 50.54(f) following Fukushima Daiichi.
1993 - 2013 Project Engineer/Manager, License Renewal and Component Replacement,
AREVA NP
Licensing Project Engineer for Prairie Island Unit 2 Replacement Steam
Generator Project, contracted in 2006 and RSGs installed in 2013. Served as
project manager beginning in 2005 for the ANO-1 EOTSG replacement
component project in charge of completing all engineering closeout activities.
Responsible for project management of Class 1 and TLAA portion of all
Entergy license renewal projects including ANO-2, PNPS, VY, Fitzpatrick,
IPEC and Cooper Nuclear Station (2002-2005). Served as the Project
Manager for the D.C. Cook, CR-3, DB-1, Pilgrim, Vermont Yankee, and ANO-
2 license renewal applications. Responsible for preparing the Reactor
Coolant System and TLAA sections of the Cook, Pilgrim, Vermont Yankee,

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Mark A. Rinckel
and ANO-2 license renewal applications.
Coordination of all working level activities for the Generic License Renewal
Program (GLRP-1992-2000). Responsibilities include coordination of FTI
activities, technical review and lead technical responsibility for all GLRP
activities, and maintenance of budget and schedule performance. Requires
significant interface with the B&W Owners Group, Industry, and the Nuclear
Regulatory Commission. License renewal activities include assessment of
aging management of critical plant equipment (e.g., Class 1 mechanical
components, non-Class 1 mechanical components, Class 1 structures, and
Electrical and I&C).
Successfully completed and received NRC approval of the following B&WOG
generic technical reports: RCS piping, pressurizer, reactor vessel, and reactor
vessel internals. The RCS piping report was the first report to receive NRC
approval in the history of license renewal. Served as the lead for the
development of the B&WOG Mechanical and Structural Tools.
Assisted Duke Energy (Oconee Units) and Entergy (ANO-1) by preparing the
technical supporting documents for the Reactor Coolant System section of the
Oconee and ANO-1 License Renewal Applications. The Oconee LRA was
approved in May 2000 and the ANO-1 application was approved in 2001.
Other license renewal activities include support of the Catawba/McGuire LRA,
review of the Class 1 Aging Management Reports for the FP&L Turkey point
Units and North Anna and Surry, and review of the NRC GALL and SRP
reports.
1990 - 1993 Supervisor, Analysis Services Unit, B&W Nuclear Technologies
Supervisor of 5-10 engineers in engineering safety analysis of the New
Production Reactor-Heavy Water Reactor Facility (NPR-HWRF). Engineers
perform all aspects of NPR safety and LOCA analysis including assessment
of advanced computer codes. Responsibilities include the preparation of
Chapter 15 for the NPR using the latest analytical tools and methods (i.e.,
RELAP5/MOD3 and best-estimate LOCA and safety analysis). Requires
significant interface with DOE, INEL, WSRC, and EBASCO.
1988 - 1990 Lead Analytical Engineer, Performance Analysis Unit, B&W Nuclear
Technologies

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Served as the technical interface between Toledo Edison and B&W with
respect to all Chapter 15 safety and licensing issues. Assignments included
resolution of LOCA limits with updated ECCS models, RPI-API statistical error
resolution, ATWS issues, Appendix R issues, steam line break and locked
rotor analyses, and TE specific MIST tests.
1984-1988 Engineer, Performance Analysis Unit, B&W Nuclear Technologies
Assignments included engineering analyses of the integrated control system,
secondary feedwater and emergency feedwater systems, operational
transients, safety and LOCA analyses, and advanced reactor designs.
Davis Besse LOFW: Performed feed and bleed analyses to resolve NRC and
UCS concerns following the June 9, 1985, LOFW event at Toledo Edison.
Transient Response of a B&W 145 FA ALWR: Prepared and delivered a
paper at the ANS conference for advanced reactors on the B&W ALWR 600
Mwe conceptual design. The paper focused on the safety features of the
design.
Code Assessments: Performed RELAP5 assessments against two integral-
effects test programs, i.e., MIST and OTIS.
177 Generic ATOG Guidelines: Performed transient analyses to support
preparation of ATOG guidelines. Assignments included RC pump restart with
voided head, asymmetric (single-loop) cooldown, and upper head cooldown
during forced and natural circulation.
Safety Analysis for the Natural Circulation Integral Reactor (NCIR):
Performed safety, LOCA and operational analyses of the conceptual design of
the NCIR ( a 40 Mwe boiling water reactor).
Design of AFW Control system for the 205-FA Plants: Participated in the
control system design of the AFW system for the 205-FA plants.
Responsibilities included analyses using the AUX-205 computer code.
1982-1984 Engineer, Fluid and Transient Analysis Unit, Nuclear Power Generation
Group
Assignments included analyses of large and small break loss-of-coolant
accidents using the suite of B&W licensing codes (i.e., CRAFT, REFLOOD,
FLECSET, THETA, and CONTEMPT). Participated on the team that resolved
NRC PTS concerns for the B&W operating plants.

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FIELD
EXPERIENCE
Steam generator tube pull/jumped steam generators (DOEL2, March 1983).
PUBLICATIONS Transient Response of a B&W 145 FA ALWR, presented at the American
Nuclear Society International Meeting on the Safety of Next Generation
Reactors, Washington, D.C., November 1990.
Benchmark of OTIS Test 2202AA With RELAP5/MOD2, BAW Report.
Demonstration of the Management of Aging Effects for Reactor Coolant
System Piping, BAW-2243A, June 1996
Demonstration of the Management of Aging Effects for the Pressurizer BAW-
2244A, August 1995
Demonstration of the Management of Aging Effects for the Reactor Vessel,
BAW-2251A, August 1999
Demonstration of the Management of Aging Effects for the Reactor Vessel
Internals, BAW-2248A, March 2000
Evaluation of Thermal Aging Embrittlement for Cast Austenitic Stainless Steel
Components, EPRI-TR-106092, Research Project 2643-33, March 1996
Reactor Pressure Vessel Integrity Program, Sixth Symposium on Current
Issues Related to Nuclear Power Plant Structures, Equipment, and Piping,
N.C. State University, December 1996
Non-Class 1 Implementation Guideline and Mechanical Tools, BAW-2270,
November 1996—Also purchased and released by EPRI.
Expert Tools to Support the Identification of Aging Effects, ASME PVP
Conference, to be presented in July 1997