ML060790179

2006/03/20-Oyster Creek, Request for Additional Information, Review of the Oyster Creek Nuclear Generating Station, License Renewal Application

ML060790179
Person / Time
Site:	Oyster Creek
Issue date:	03/20/2006
From:	Ashley D NRC/NRR/ADRO/DLR/RLRA
To:	Swenson C AmerGen Energy Co
Ashley D, NRR/DLR/RLRA, 415-3191
References
%dam200606, TAC MC7624
Download: ML060790179 (9)

Text

March 20, 2006C. N. SwensonSite Vice President AmerGen Energy Company, LLC

P.O. Box 388 Forked River, NJ 08731

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THEOYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWALAPPLICATION (TAC NO. MC7624)

Dear Mr. Swenson:

By letter dated July 22, 2005, AmerGen Energy Company, LLC (AmerGen or the applicant)submitted to the U.S. Nuclear Regulatory Commission (NRC or the staff) an applicationpursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54), to renew theoperating license for Oyster Creek Nuclear Generating Station. The NRC staff is reviewing theinformation contained in the license renewal application and has identified, in the enclosure,areas where additional information is needed to complete the review.These questions were discussed with members of your staff during a conference call onFebruary 8, 2006. A mutually agreeable date for a response is within 30 days from the date ofthis letter. If you have any questions, please contact me at 301-415-3191 or via e-mail at DJA1@nrc.gov

.Sincerely,/RA/ Donnie J. Ashley, Project ManagerLicense Renewal Branch A Division of License Renewal Office of Nuclear Reactor RegulationDocket No. 50-219

Enclosure:

As statedcc w/encl: See next page

DOCUMENT NAME: E:\Filenet\ML060790179.wpdOFFICEPM:RLRALA:DLRBC:RLRANAMEDAshleyYEdmondsLLund DATE03/ 7 /0603/ 16 /0603/ 20 /06 Oyster Creek Nuclear Generating Station 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 Letter to C. N. Swenson from Donnie J. Ashley dated March 20, 2006

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THEOYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWALAPPLICATION (TAC NO. MC7624)HARD COPYDLR R/FE-MAIL:JFairRWeisman AMurphy RPettis GGalletti CLi GBagchi SSmith (srs3)

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RLauferGMiller RBellamy, RI RCureton, RI JLilliendahl, RIMModes, RI MSykes, RI AHodgdon DShum OPA RidsNrrDorl OYSTER CREEK NUCLEAR GENERATING STATIONLICENSE RENEWAL APPLICATION (LRA) REQUEST FOR ADDITIONAL INFORMATION (RAI)RAI 2.4.1-1A review of LRA Table 2.4.1 indicates that drywell seismic supports and anchorages are notwithin the scope of license renewal, though they are relied upon for drywell stability. Acomponent type "Biological Shield Wall - Lateral Support" is in the Table. The staff requests the applicant to provide justification for not including the drywell seismic lateral supports and anchorages within the scope of license renewal.RAI 2.4.1-2LRA Tables 2.4.1 and 2.4.2 do not incorporate refueling cavity seal components within thescope of license renewal, although the plant has experienced significant corrosion (as described in Item 4 of LRA Section 3.5.2.2) of the drywell as a result of leakage from the seal.

The staff requests the applicant to include the seal in the scope of license renewal, or provide justification for not including it within the scope of license renewal.RAI 2.4.2-1LRA Page 2.4.8 states that structural seals are within the boundary of evaluation, but withoutexplaining what they are. The staff requests the applicant to identify all the structural seals inthe reactor building.RAI 2.4.8-1LRA Section 2.4.8, Fire Pond dam, states that the dam is classified as Safety Class III. Thestaff requests the applicant to identify the location in the LRA or updated final safety analysis report (UFSAR) where the definition of Safety Class III is provided. If the definition was notprovided in the LRA or UFSAR, the staff requests the applicant to provide a definition for Safety Class III.

RAI 2.4.9-1LRA Section 2.4.9, Fire Pumphouses, states that the pumphouse and the tank foundations areclassified non-safety related, Seismic Class II. The staff request the applicant to identify the location in the LRA or UFSAR where the definition of "non-safety related, Seismic Class II" is provided. If the definition was not provided in the LRA or UFSAR, the staff requests the applicant to provide a definition for "non-safety related, Seismic Class II."

RAI 3.5-1LRA Table 3.5.2.1.1 indicates that fretting and lockup of suppression pool downcomers will bemanaged by ASME Section XI, Subsection IWE (AMP B.1.27). Directly, the downcomers are not part of the pressure boundary. Subsection IWE does not provide examination requirementsand acceptance criteria for downcomers. However, as a convenience, the examinations of downcomers can be included in SubsectionIWE requirements, with special provisions for examining the downcomers for fretting or lockups in the plant-specific procedures. The staff requests the applicant to provide (1) a discussion of operating experience related to downcomers fretting or lockups, and (2) the ISI provisions incorporated in the plant-specific IWE program. RAI 3.5-2LRA Table 3.5.2.1.1 credits 10 CFR Part 50, Appendix J (AMP B.1.29) for management ofdowncomers "Loss of Material." It is not apparent, how the leak testing requirement of Appendix J will detect loss of material of downcomers. The staff requests the applicant todiscuss the use of Appendix J in managing loss of material in downcomers.RAI 3.5-3Under component types "Reactor Pedestal" and "R.C. Floor Slab," a reference is made to LRA Table 1 Item 3.5.1-29. The discussion in Item 3.5.1-29 indicates that the concretetemperatures in the upper part of the drywell could be as high as 259F. As a result, thereactor building drywell shield concrete had significant cracking. However, the cause of the high temperature is not indicated. In light of the above discussion, the staff requests theapplicant to provide the following information:a.Type and adequacy of the cooling system used to control the temperatures in the drywell.b.Operating experience related to the reliability of the cooling system.

c.Actions taken to reduce the high temperatures in the upper part of the drywell.

d.A summary of the results of the last inspection of reactor pedestal, R.C. floor slabs,drywell lateral supports, and sacrificial shield wall, including the date of the inspection, and frequencies of inspection during the period of extended operation.

RAI 3.5-4Component type "Shielding Blocks and Plates," uses patented material "Permali," for which noaging effects are indicated in LRA Table 3.5.2.1.1. The staff requests the applicant to provide abrief description of the material, and the AMR results that justified that it does not need agingmanagement during the period of extended operation.RAI 3.5-5For all component types described in LRA Table 3.5.2.1.1 (Primary Containment), the "waterchemistry program" is vital for the components fully or partially submerged in water, in addition to the programs noted in the individual component types. The staff requests the applicant to provide reasons for not including the water chemistry program to manage the aging degradation of these components. RAI 3.5-6The through-wall cracking of the Fitzpatrick torus indicates a need for closer examination of thehighly restrained and structurally discontinuous areas subject to operational cyclic loads. Theprime aging management program used for managing degradation of the primary containment structure is Subsection IWE (AMP B.1.27). The program is focused towards detecting loss of material. The staff requests the applicant to discuss how the program would detect initiation of such cracking in the Oyster Creek primary containment.RAI 3.5-7LRA Table 3.5.3.1.18 indicates that the aging of Class MC component supports is managed byASME Section XI, Subsection IWF during the CLB. However, a review of the "Enhancement" in AMP B.1.28 (ASME Section XI, Subsection IWF) indicates that the program will be enhancedduring the period of extended operation to include additional MC supports and underwater structures in the torus. The staff requests the applicant to provide clarifications regarding the inspection of Class MC supports during the CLB and during the period of extended operation. RAI 3.5-8LRA Tables 3.5.2.1.6, 3.5.2.

1.15, 3.5.2.1.16, and 3.5.2.1.17 identify loss of preload as theaging effect requiring management for structural bolts, and the structural monitoring program (B.1.31) as its aging management program. The Structural Monitoring Program states that exposed surfaces of bolting are monitored for indications of loss of preload, and that the program relies on procurement controls and installation practices, defined in plant procedures, to ensure that only approved lubricants and proper torque are applied consistent with the GALLReport bolting integrity program. LRA B.1.12 Bolting Integrity Program states that the programtakes exception to the GALL Report and that the aging management of structural bolting isaddressed by the Structural Monitoring Program. The staff requests the applicant to address the following: a.The applicant needs to resolve the apparent inconsistency that the StructuralMonitoring Program states that the proper torque for bolts is applied consistent with the GALL Report bolting integrity program while the Bolting Integrity Program takes exception to the GALL Report and refers the aging management of structural bolting back to the structural monitoring program.b.Does the applicant identify loss of preload of structural bolts by visual inspection orby applying a torque wrench? If it is by visual inspection, explain how the loss of preload can be estimated by visual inspection.c.LRA Section B.1.31 states that the Structural Monitoring Program relies onprocurement controls and installation practices, defined in plant procedures, to ensure that only approved lubricants and proper torque are applied. The staff believes that bolt procurement controls and installation practices were supposedly used before, during, or immediately after the bolts were installed. Since the Structural Monitoring Program is being used to inspect structural bolts after the bolts were installed for sometime, the staff requests the applicant to explain how could the Structural Monitoring Program rely on bolt procurement controls andinstallation practices.d.Are there any structural bolts or fasteners, which have a yield strength equal to orgreater than 150 ksi, managed by the structural monitoring program? If yes, provide justification for not using the bolting integrity program as the aging management program for structural bolts. RAI 3.5-9LRA Table 3.5.2.1.7 lists the Structural Monitoring Program as the AMP for penetration seals ofelastomer and grout in the soil environment. The AMP in LRA, Appendix B states that the program will require inspection of penetration seals, but does not state how the inspectionshould be conducted for penetration seals of elastomer and grout in the soil environment and the frequency of the inspection. The staff requests the applicant to describe the inspection method and frequency for penetration seals of elastomer and grout in the soil environment. RAI 3.5-10LRA Table 3.5.2.1.7 lists aluminum material embedded in concrete, and states no aging effectand requiring no AMP. The ACI Building Code prohibits the use of aluminum in structural concrete unless it is coated or covered to prevent aluminum-concrete reaction or electrolytic action between aluminum and steel. The staff requests the applicant to justify the use of aluminum material in concrete and to explain why there is no aging effect and that an AMP is not required.