ML060890660

Request for Additional Information for the Review of the Oyster Creek Nuclear Generating Station, License Renewal Application - Application Sections 4.2 and 4.7 (TAC No. MC76240)

ML060890660
Person / Time
Site:	Oyster Creek
Issue date:	03/30/2006
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: ML060890660 (7)

Text

March 30, 2006 C. N. Swenson Site Vice President AmerGen Energy Company, LLC P.O. Box 388 Forked River, NJ 08731-0388

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL APPLICATION (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 application pursuant to Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54), to renew the operating license for Oyster Creek Nuclear Generating Station. The NRC staff is reviewing the information 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 on March 16, 2006. A mutually agreeable date for a response is within 30 days from the date of this 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 Manager License Renewal Branch A Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-219

Enclosure:

As stated cc w/encl: See next page

DOCUMENT NAME: E:\\Filenet\\ML060890660.wpd OFFICE PM:RLRA LA:DLR BC:RLRA NAME DAshley YEdmonds LLund DATE 03/27/06 03/24/06 03/30/06

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 08731 Senior Vice President of Operations AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348 Kathryn M. Sutton, Esquire Morgan, Lewis, & Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Kent Tosch, Chief New Jersey Department of Environmental Protection Bureau of Nuclear Engineering CN 415 Trenton, NJ 08625 Vice President - Licensing and Regulatory Affairs AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 445 Forked River, NJ 08731 Director - Licensing and Regulatory Affairs AmerGen Energy Company, LLC Correspondence Control P.O. Box 160 Kennett Square, PA 19348 Manager Licensing - Oyster Creek Exelon Generation Company, LLC Correspondence Control P.O. Box 160 Kennett Square, PA 19348 Regulatory Assurance Manager Oyster Creek AmerGen Energy Company, LLC P.O. Box 388 Forked River, NJ 08731 Assistant General Counsel AmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348 Ron Bellamy, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Correspondence Control Desk AmerGen Energy Company, LLC 200 Exelon Way, KSA 1--1 Kennett Square, PA 19348 Oyster Creek Nuclear Generating Station Plant Manager AmerGen Energy Company, LLC P.O. Box 388 Forked River, NJ 08731 License Renewal Manager Exelon Generation Company, LLC 200 Exelon Way, Suite 230 Kennett Square, PA 19348

Oyster Creek Nuclear Generating Station cc:

Mr. James Ross Nuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 Mr. Michael P. Gallagher Vice President License Renewal Exelon Generation Company, LLC 200 Exelon Way, Suite 230 Kennett Square, PA 19348 Mr. Christopher M. Crane President and Chief Nuclear Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555

Ltr. to C.N. Swenson from Donnie Ashley dated: March 30, 2006

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL APPLICATION (TAC NO. MC7624)

Adams Accession No.: ML060890660 HARD COPY DLR R/F DISTRIBUTION:

JFair RWeisman AMurphy RPettis GGalletti CLi GBagchi SSmith (srs3)

SDuraiswamy YL (Renee) Li RidsNrrDlr RidsNrrDlrRlra RidsNrrDlrRlrb RidsNrrDe RidsNrrDci RidsNrreEemb RidsNrrDeEeeb RidsNrrDeEqva RidsNrrDss RidsNrrDnrl RidsOgcMailCenter RidsNrrAdes DLR Staff VRodriguez DAshley RLaufer GMiller RBellamy, RI RCureton, RI JLilliendahl, RI MModes, RI MSykes, RI AHodgdon DShum RidsOpaMail RidsNrrDorl

OYSTER CREEK NUCLEAR GENERATING STATION LICENSE RENEWAL APPLICATION (LRA)

REQUEST FOR ADDITIONAL INFORMATION (RAI)

RAI 4.2.2-1 Please provide the bounding values for the percentage decrease in the upper-shelf energy (USE) at 50 effective full-power years (EFPY) based on the equivalent margin analysis (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.

RAI 4.2.2-2 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 staffs reactor vessel integrity database (RVID) for these welds.

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 4.0 are valid for the above welds.

b. justification for the use of these chemistry data for the above welds, including the source of the data, and a specific reference for the documentation/analysis demonstrating that these chemistry data represent the best available estimate of the weld chemistries.

RAI 4.2.7-1 Boiling Water Reactor (BWR) Vessels and Internals Project-26, BWR Vessels and Internals Project, BWR Top Guide Inspection and Flaw Evaluation Guidelines, indicates that BWR stainless steel components exposed to a fluence greater than 5 x 1020 n/cm2 (E > 1 MeV) are susceptible 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.

Section 4.2.7, Reactor 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 1020/cm2 (E > 1 MeV) and are, therefore, considered susceptible to IASCC, based on fluence calculations performed for these components. However, no TLAA associated with IASCC exists for the core shroud, incore instrumentation dry tubes, or top guide.

Please clarify why there is no TLAA for the core shroud, incore instrumentation dry tubes, and top guide given that these components have been exposed to a fluence exceeding 5 x 1020 n/cm2 (E > 1 MeV) and are considered susceptible to IASCC.

RAI 4.7.4-1 Section 4.7.4, Reactor Vessel Weld Flaw Evaluations, of the LRA includes a discussion of several 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 and found 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 with the ASME Code,Section XI, IWB-3600 for conditions of extended operation through 50 EFPY.

RAI 4.7.5-1 Section 4.7.5, Control Rod Drive (CRD) Stub Tube Flaw Analysis, of the LRA includes a discussion of several cracks that were found in the CRD stub tubes during construction and a subsequent repair of the cracks. This repair includes the grinding out of the observed cracks followed by the application 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 the repairs 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 shutdown cycles would be required in order for any such postulated crack to propagate through the overlay to the surface of the CRD stub tube. The LRA states that this information was provided to 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 of extended operation is projected to be less than 275. Therefore, the LRA stipulates that the above evaluation remains valid for ensuring CRD stub tube integrity through the end of the period of extended operation.

Given the extent of operating experience since the time of the original analysis, there is a possibility that other CRD stub tube degradation mechanisms that were not known or considered at the time of the original analysis could potentially compromise the integrity of the CRD stub tube over the period of extended operation. Please discuss whether there are any known degradation mechanisms discovered since the time of implementation of the CRD stub tube repair that could potentially invalidate the original analysis discussed above. If any CRD stub tube degradation mechanism is known to exist that was not taken into consideration at the time of the original analysis, thereby, potentially invalidating that analysis, please submit a revised TLAA for the CRD stub tubes demonstrating that the integrity of these components will be maintained over the period of extended operation.