ML062230429

2006/08/11-Summary of a Telephone Conference Call Held on March 16, 2006, Between the NRC and Amergen

ML062230429
Person / Time
Site:	Oyster Creek
Issue date:	08/11/2006
From:	Ashley D NRC/NRR/ADRO/DLR/RLRA
To:	AmerGen Energy Co
Ashley D, NRR/DLR/RLRA, 415-3191
References
%dam200612
Download: ML062230429 (9)

Text

August 11,2006LICENSEE:AmerGen Energy Company, LLC FACILITY:Oyster Creek Nuclear Generating Station

SUMMARY

OF A TELEPHONE CONFERENCE CALL HELD ON MARCH 16, 2006, BETWEEN THE U.S. NUCLEAR REGULATORYCOMMISSION AND AMERGEN ENERGY COMPANY, LLC, CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWALAPPLICATIONThe U.S. Nuclear Regulatory Commission staff (NRC or the staff), and representatives ofAmerGen Energy Company, LLC (AmerGen), held a telephone conference call on March 16, 2006, to discuss and clarify the staff's draft requests for additional information (D-RAIs) concerning the Oyster Creek Nuclear Generating Station license renewal application. The conference call was useful in clarifying the intent of the staff's D-RAIs.Enclosure 1 provides a listing of the conference call participants. Enclosure 2 contains a listingof the D-RAI discussed with the applicant, including a brief description on the status of the items.The applicant had an opportunity to comment on this summary./RA/Donnie J. Ashley, Project ManagerLicense Renewal Branch A Division of License Renewal Office of Nuclear Reactor RegulationDocket No. 50-219

Enclosures:

As statedcc w/encls: See next page

ML062230429 DOCUMENT NAME: E:\Filenet\ML062230429.wpdOFFICEPM:RLRA:DLRLA:RLRA:DLRBC:RLRA:DLRNAMEDAshley YEdmonds LLundDATE08/ 11 /0608/ 10 /0608/ 11 /06 cc:Site Vice President - Oyster Creek Nuclear Generating Station AmerGen Energy Company, LLC

P.O. Box 388 Forked River, NJ 08731Senior Vice President of Operations AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348Kathryn M. Sutton, EsquireMorgan, Lewis, & Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004Kent Tosch, ChiefNew Jersey Department of Environmental Protection Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625Vice President - Licensing and Regulatory Affairs AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555Regional Administrator, Region IU.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415Mayor of Lacey Township818 West Lacey Road Forked River, NJ 08731Senior Resident InspectorU.S. Nuclear Regulatory Commission

P.O. Box 445 Forked River, NJ 08731Director - Licensing and Regulatory AffairsAmerGen Energy Company, LLC Correspondence Control

P.O. Box 160 Kennett Square, PA 19348Manager Licensing - Oyster CreekExelon Generation Company, LLC Correspondence Control

P.O. Box 160 Kennett Square, PA 19348Regulatory Assurance Manager Oyster Creek AmerGen Energy Company, LLC

P.O. Box 388 Forked River, NJ 08731Assistant General CounselAmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348Ron Bellamy, Region IU.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415Correspondence Control DeskAmerGen Energy Company, LLC 200 Exelon Way, KSA 1--1 Kennett Square, PA 19348Oyster Creek Nuclear Generating StationPlant Manager AmerGen Energy Company, LLC

P.O. Box 388 Forked River, NJ 08731License Renewal ManagerExelon Generation Company, LLC 200 Exelon Way, Suite 230 Kennett Square, PA 19348 Oyster Creek Nuclear Generating Station cc:

Mr. James RossNuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708Mr. Michael P. GallagherVice President License Renewal Exelon Generation Company, LLC 200 Exelon Way, Suite 230 Kennett Square, PA 19348Mr. Christopher M. CranePresident and Chief Nuclear Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 Note to: AmerGen Energy Company, LLC, Facility: Oyster Creek Nuclear Generating Stationfrom Donnie Ashley dated August 11, 2006.

SUMMARY

OF A TELEPHONE CONFERENCE CALL HELD ON MARCH 16, 2006, BETWEEN THE U.S. NUCLEAR REGULATORYCOMMISSION AND AMERGEN ENERGY COMPANY, LLC, CONCERNING DRAFT REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWALAPPLICATIONDISTRIBUTION

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DAshley VRodriguez RLaufer GMiller RBellamy, RI RCureton, RI JLilliendahl, RIMModes, RI MSykes, RI AHodgdon RidsOpaMail ENCLOSURE 1LIST OF PARTICIPANTS FOR TELEPHONE CONFERENCE CALLTO DISCUSS THE OYSTER CREEK NUCLEAR GENERATING STATIONLICENSE RENEWAL APPLICATIONMarch 16, 2006Participants AffiliationsDonnie AshleyU.S. Nuclear Regulatory Commission (NRC)Matt MitchellNRC Chris SydnorNRC Ganesh CheruvenkiNRC George BeckAmerGen Energy Company, LLC (AmerGen)

Mike MayAmerGen Gary StevensAmerGen Greg HarttraftAmerGen ENCLOSURE 2DRAFT REQUESTS FOR ADDITIONAL INFORMATION (D-RAIs)OYSTER CREEK NUCLEAR GENERATING STATIONLICENSE RENEWAL APPLICATIONMarch 16, 2006The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of AmerGenEnergy Company, LLC (AmerGen), held a telephone conference call on March 16, 2006, to discuss and clarify the staff's draft requests for additional information (D-RAIs) concerning the Oyster Creek Nuclear Generating Station, license renewal application (LRA). The following D-RAIs were discussed during the telephone conference call.D-RAI 4.2.2-1Please provide the bounding values for the percentage decrease in the upper-shelf energy (USE)at 50 effective full-power year (EFPY) based on the EMA, as well as the predicted percentage decrease in USE at 50 EFPY, as determined from Regulatory Guide (RG) 1.99, Revision 2, for all Reactor Vessel (RV) beltline materials.Discussion:The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI.D-RAI 4.2.2-2In LRA Table 4.2.2-1 in Section 4.0 of the LRA provides the Adjusted Reference Temperature(ART) values for the RV beltline materials. The chemistry data (%Cu and %Ni) and chemistry factor (CF) values for the Lower-to-Lower Intermediate Shell Circumferential Weld 3-564; Lower Shell Axial Welds 2-564A, B, and C; and Lower Intermediate Shell Axial Welds 2-564D, E, and F from Table 4.2.2-1 are less conservative than the corresponding chemistry data and CF values that were established in the staff's reactor vessel integrity database (RVID) for thesewelds. Please supplement Section 4.0 of the LRA with the following information:a.Verification of whether the chemistry data contained in Table 4.2.2-1 of Section 4.0are valid for the above welds. b. Justification for the use of these chemistry data for the above welds, including thesource of the data, and a specific reference for the documentation/analysisdemonstrating that these chemistry data represent the best available estimate of theweld chemistries.Discussion:The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI. D-RAI 4.2.7-1Boiling Water Reactor Vessel and Internals Project (BWRVIP)-26, "BWR Vessels and InternalsProject, BWR Top Guide Inspection and Flaw Evaluation Guidelines," indicates that BWR stainless steel components exposed to a fluence greater than 5 x 10 20 n/cm 2 (E > 1 MeV) aresusceptible to irradiation assisted stress corrosion cracking (IASCC). The safety evaluation report (SER) for BWRVIP-26 considers IASCC of BWR reactor internals a time-limited aging analysis (TLAA) issue.In Section 4.2.7, Reactors Internals Components, of the LRA indicates that the core shroud,incore instrumentation dry tubes, and top guide have been exposed to a fluence exceeding 5 x 10 20 n/cm 2 (E > 1 MeV) and are, therefore, considered susceptible to IASCC, based onfluence calculations performed for these components. However, no TLAA associated with IASCC exists for the core shroud, incore instrumentation drytubes, or top guide.Please clarify why there is no TLAA for the core shroud, incore instrumentation drytubes, andtop guide given that these components have been exposed to a fluence exceeding 5 x 10 20 n/cm 2 (E > 1 MeV) and are considered susceptible to IASCC.Discussion:The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI.D-RAI 4.7.4-1In Section 4.7.4, of the LRA Reactor Vessel Weld Flaw Evaluations, includes a discussion ofseveral flaws that were detected in two axial RV welds during inservice inspections performed in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME) Code,Section XI and documented in a 2000 Inservice Inspection Report. The flaws were previously evaluated and found to be acceptable for the current licensing period in accordance with the ASME Code,Section XI, IWB-3600. Section 4.7.4 of the LRA indicates that these flaws were reevaluated for conditions of extended operation through 50 EFPY andfound to be acceptable in accordance with the American Society of Mechanical Engineers (ASME) Code,Section XI, IWB-3600.Please submit the analysis demonstrating that these flaws are acceptable in accordance withthe ASME Code,Section XI, IWB-3600 for conditions of extended operation through 50 EFPY.Discussion:The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI.D-RAI 4.7.5-1In Section 4.7.5, of the LRA Control Rod Drive (CRD) Stub Tube Flaw Analysis, includes adiscussion of several cracks that were found in the CRD stub tubes during construction and asubsequent repair of the cracks that included grinding out the observed cracks followed by theapplication of a weld overlay. The LRA indicated that, following the repair, an analysis was performed to demonstrate that any crack that would have remained undetected following therepairs would not propagate through the weld overlay during the life of the plant. Furthermore, the LRA states that this analysis demonstrated that more than 1000 startup and shutdowncycles would be required in order for any such postulated crack to propagate through theoverlay to the surface of the CRD stub tube. The LRA states that this information was providedto the Atomic Energy Commission (AEC) in Amendment 37 to the provisional operating license application.The cumulative number of startup and shutdown cycles through the end of the period ofextended operation is projected to be less than 275. Therefore, the LRA stipulates that theabove evaluation remains valid for ensuring CRD stub tube integrity through the end of theperiod of extended operation. Given the extent of operating experience since the time of the original analysis, there is apossibility that other CRD stub tube degradation mechanisms that were not known orconsidered at the time of the original analysis could potentially compromise the integrity of theCRD stub tube over the period of extended operation. Please discuss whether there are anyknown degradation mechanisms discovered since the time of implementation of the CRD stubtube repair that could potentially invalidate the original analysis discussed above. If any CRDstub tube degradation mechanism is known to exist that was not taken into consideration at thetime of the original analysis, thereby, potentially invalidating that analysis, please submit arevised TLAA for the CRD stub tubes demonstrating that the integrity of these components willbe maintained over the period of extended operation.Discussion:The applicant indicated that the question is clear. This D-RAI will be sent as aformal RAI.