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| | number = ML062230429 | | | number = ML062230429 |
| | issue date = 08/11/2006 | | | issue date = 08/11/2006 |
| | title = 2006/08/11-Summary of a Telephone Conference Call Held on March 16, 2006, Between the NRC and Amergen | | | title = Summary of a Telephone Conference Call Held on March 16, 2006, Between the NRC and Amergen |
| | author name = Ashley D | | | author name = Ashley D |
| | author affiliation = NRC/NRR/ADRO/DLR/RLRA | | | author affiliation = NRC/NRR/ADRO/DLR/RLRA |
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| =Text= | | =Text= |
| {{#Wiki_filter:August 11,2006LICENSEE:AmerGen Energy Company, LLC FACILITY:Oyster Creek Nuclear Generating Station | | {{#Wiki_filter:August 11,2006 LICENSEE: AmerGen Energy Company, LLC FACILITY: Oyster Creek Nuclear Generating Station |
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| ==SUBJECT:== | | ==SUBJECT:== |
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| ==SUMMARY== | | ==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 | | OF A TELEPHONE CONFERENCE CALL HELD ON MARCH 16, 2006, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND AMERGEN ENERGY COMPANY, LLC, CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL APPLICATION The U.S. Nuclear Regulatory Commission staff (NRC or the staff), and representatives of AmerGen Energy Company, LLC (AmerGen), held a telephone conference call on March 16, 2006, to discuss and clarify the staffs 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 staffs D-RAIs. provides a listing of the conference call participants. Enclosure 2 contains a listing of 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 Manager License Renewal Branch A Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-219 |
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| ==Enclosures:== | | ==Enclosures:== |
| As statedcc w/encls: See next page
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| 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
| | As stated cc w/encls: See next page |
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| 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 | | ML062230429 DOCUMENT NAME: E:\Filenet\ML062230429.wpd OFFICE PM:RLRA:DLR LA:RLRA:DLR BC:RLRA:DLR NAME DAshley YEdmonds LLund DATE 08/ 11 /06 08/ 10 /06 08/ 11 /06 cc: |
| | Site Vice President - Oyster Creek Director - Licensing and Regulatory Affairs Nuclear Generating Station AmerGen Energy Company, LLC AmerGen Energy Company, LLC Correspondence Control P.O. Box 388 P.O. Box 160 Forked River, NJ 08731 Kennett Square, PA 19348 Senior Vice President of Manager Licensing - Oyster Creek Operations Exelon Generation Company, LLC AmerGen Energy Company, LLC Correspondence Control 200 Exelon Way, KSA 3-N P.O. Box 160 Kennett Square, PA 19348 Kennett Square, PA 19348 Kathryn M. Sutton, Esquire Regulatory Assurance Manager Morgan, Lewis, & Bockius LLP Oyster Creek 1111 Pennsylvania Avenue, NW AmerGen Energy Company, LLC Washington, DC 20004 P.O. Box 388 Forked River, NJ 08731 Kent Tosch, Chief New Jersey Department of Assistant General Counsel Environmental Protection AmerGen Energy Company, LLC Bureau of Nuclear Engineering 200 Exelon Way CN 415 Kennett Square, PA 19348 Trenton, NJ 08625 Ron Bellamy, Region I Vice President - Licensing and U.S. Nuclear Regulatory Commission Regulatory Affairs 475 Allendale Road AmerGen Energy Company, LLC King of Prussia, PA 19406-1415 4300 Winfield Road Warrenville, IL 60555 Correspondence Control Desk AmerGen Energy Company, LLC Regional Administrator, Region I 200 Exelon Way, KSA 1--1 U.S. Nuclear Regulatory Commission Kennett Square, PA 19348 475 Allendale Road King of Prussia, PA 19406-1415 Oyster Creek Nuclear Generating Station Plant Manager Mayor of Lacey Township AmerGen Energy Company, LLC 818 West Lacey Road P.O. Box 388 Forked River, NJ 08731 Forked River, NJ 08731 Senior Resident Inspector License Renewal Manager U.S. Nuclear Regulatory Commission Exelon Generation Company, LLC P.O. Box 445 200 Exelon Way, Suite 230 Forked River, NJ 08731 Kennett Square, PA 19348 |
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| P.O. Box 445 Forked River, NJ 08731Director - Licensing and Regulatory AffairsAmerGen Energy Company, LLC Correspondence Control | | 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 |
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| P.O. Box 160 Kennett Square, PA 19348Manager Licensing - Oyster CreekExelon Generation Company, LLC Correspondence Control
| | Note to: AmerGen Energy Company, LLC, Facility: Oyster Creek Nuclear Generating Station from Donnie Ashley dated August 11, 2006. |
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| P.O. Box 160 Kennett Square, PA 19348Regulatory Assurance Manager Oyster Creek AmerGen Energy Company, LLC
| | ==SUBJECT:== |
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| 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
| | ==SUMMARY== |
| | OF A TELEPHONE CONFERENCE CALL HELD ON MARCH 16, 2006, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND AMERGEN ENERGY COMPANY, LLC, CONCERNING DRAFT REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL APPLICATION DISTRIBUTION: |
| | HARD COPY DLR R/F E-MAIL: |
| | 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 DAshley VRodriguez RLaufer GMiller RBellamy, RI RCureton, RI JLilliendahl, RI MModes, RI MSykes, RI AHodgdon RidsOpaMail |
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| 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:
| | LIST OF PARTICIPANTS FOR TELEPHONE CONFERENCE CALL TO DISCUSS THE OYSTER CREEK NUCLEAR GENERATING STATION LICENSE RENEWAL APPLICATION March 16, 2006 Participants Affiliations Donnie Ashley U.S. Nuclear Regulatory Commission (NRC) |
| 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.
| | Matt Mitchell NRC Chris Sydnor NRC Ganesh Cheruvenki NRC George Beck AmerGen Energy Company, LLC (AmerGen) |
| | Mike May AmerGen Gary Stevens AmerGen Greg Harttraft AmerGen ENCLOSURE 1 |
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| ==SUBJECT:==
| | DRAFT REQUESTS FOR ADDITIONAL INFORMATION (D-RAIs) |
| | OYSTER CREEK NUCLEAR GENERATING STATION LICENSE RENEWAL APPLICATION March 16, 2006 The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of AmerGen Energy Company, LLC (AmerGen), held a telephone conference call on March 16, 2006, to discuss and clarify the staffs 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 4.2.2-1 Please 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 a formal RAI. |
| | D-RAI 4.2.2-2 In 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 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. |
| | Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI. |
| | ENCLOSURE 2 |
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| ==SUMMARY==
| | D-RAI 4.2.7-1 Boiling Water Reactor Vessel and Internals Project (BWRVIP)-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. |
| 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
| | 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 1020 n/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 drytubes, or top guide. |
| :HARD COPYDLR R/FE-MAIL:JFairRWeisman AMurphy RPettis GGalletti CLi GBagchi SSmith (srs3)
| | Please clarify why there is no TLAA for the core shroud, incore instrumentation drytubes, 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. |
| SDuraiswamy YL (Renee) Li RidsNrrDlr RidsNrrDlrRlra RidsNrrDlrRlrb RidsNrrDe RidsNrrDci RidsNrreEemb RidsNrrDeEeeb RidsNrrDeEqva RidsNrrDss RidsNrrDnrl RidsOgcMailCenter RidsNrrAdes DLR Staff
| | Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI. |
| -------------
| | D-RAI 4.7.4-1 In Section 4.7.4, of the LRA Reactor Vessel Weld Flaw Evaluations, 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. |
| 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)
| | 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. |
| 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.}}
| | Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI. |
| | D-RAI 4.7.5-1 In Section 4.7.5, of the LRA Control Rod Drive (CRD) Stub Tube Flaw Analysis, includes a discussion of several cracks that were found in the CRD stub tubes during construction and a subsequent repair of the cracks that included grinding out 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. |
| | Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI. |
| | }} |
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Category:Meeting Summary
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Meeting Summary with the Beyond Nuclear Petitioners Regarding Their 2.206 Petition to Suspend Operating Licenses (Ols) of General Electric (GE) Boiling Water Reactors (Bwrs) Mark I (Bwrs)
ML1036305532011-01-0606 January 2011
Summary of Meeting (Teleconference) with Exelon to Discuss RAI Regarding the Implementation of an Emergency Action Level Scheme Based on NEI 99-01, Rev. 5
ML1020700252010-07-27027 July 2010
Summary of Category 1 Meeting with Exelon to Discuss Re-Submittal of Its Cyber Security Plan
ML1015401502010-06-0202 June 2010
Summary of Annual Assessment Meeting, for Oyster Creek
ML0930804802009-11-19019 November 2009
Meeting Summary of 10/22/2009 Meeting Regarding Its Power Uprate Project Plan
ML0930705002009-11-0505 November 2009
Summary of Meeting with Exelon to Discuss Oyster Creek Nuclear Generating Station and Exelon Fleet Lessons-Learned on Buried Piping Issues
ML0927400612009-10-22022 October 2009
Summary of Meeting with Exelon to Discuss Proposed License Amendment Regarding Secondary Containment Operability
ML0919803632009-08-20020 August 2009
Summary of Licensing Counterparts Meeting with Exelon Generation Company, to Discuss Licensing Practices and Processes
ML0915301992009-06-0202 June 2009
Summary of Annual Assessment Public Meeting, 5/28/09, with Sign-in Sheets
ML0920200532008-12-0505 December 2008
Drop-in Visit- December 5, 2008, Rev 2
ML0833100732008-11-26026 November 2008
Summary of Meeting with Exelon to Review the Results of Security Organization Safety Conscious Work Environment (SCWE) Surveys at Exelon Nuclear Operating Reactors
ML0812007572008-06-0202 June 2008
Summary of Category 1 Meeting with Exelon to Discuss Current and Planned Licensing Activities and Processes
ML0814105622008-05-20020 May 2008
Summary of Annual Assessment Meeting with Amergen Energy Company, LLC
ML0802503822008-01-23023 January 2008
Summary - Drop-in Visit from Exelon/Amergen, 1/15/08
ML0709303792007-06-11011 June 2007
Summary of Category 1 Meeting with Exelon to Discuss Current and Planned Licensing Activities and Processes to Discuss Topics of Common Interest to the Reactor Licensing Programs for Exelon and the NRC Staff
2024-01-23
[Table view]
Category:Request for Additional Information (RAI)
MONTHYEARML21099A0442021-04-0808 April 2021
Enclosure - Request for Additional Information Related to Request for Approval of Oyster Creek Nuclear Generating Station Independent Spent Fuel Storage Installation Only Emergency Plan Holtec Decommissioning International, LLC
ML20335A3102020-11-30030 November 2020
RAI November 2020 Part 73 FOF Exemption Request
ML20297A2382020-10-22022 October 2020
Fleet DQAP RAIs - Enclosure
ML20297A2372020-10-22022 October 2020
Request for Additional Information - HDI Fleet Decommissioning Quality Assurance Program
ML20134H8732020-05-12012 May 2020
E-mail - Request for Additional Information: Oyster Creek Request for Temporary Exemption from Part 73, Appendix B Requirements
ML19263D2142019-09-20020 September 2019
Hdi PSDAR Environmental RAI Enclosure (002)
ML19158A2772019-05-28028 May 2019
NRR E-mail Capture - Oyster Creek ESA Consultation: Information Requests from NMFS
ML19025A1172019-01-24024 January 2019
NRR E-mail Capture - for Your Action - RAI for Oyster Creek EP Exemption (Change to Adiabatic Heat-Up Calculation)
ML18256A0022018-09-12012 September 2018
NRR E-mail Capture - for Your Action: RAIs for Oyster Creek Post Shutdown Decommissioning Activities Report (PSDAR)
ML18135A0732018-05-15015 May 2018
NRR E-mail Capture - RAIs - Oyster Creek - Request Approval of the Decommissioning Quality Assurance Program
ML18068A6582018-03-0909 March 2018
NRR E-mail Capture - RAIs - Oyster Creek Defueled TS LAR
ML18067A1482018-03-0808 March 2018
Enclosurequest for Additional Information (Letter to P. R. Simpson Request for Additional Information Regarding Exelon Generating Company, Llc'S Decommissioning Funding Plan Update for Independent Spent Fuel Storage Installation)
ML18024B0642018-01-24024 January 2018
NRR E-mail Capture - for Your Action - RAI for Oyster Creek Permanently Defueled EP and EAL Scheme LAR (CAC No. MG0160; EPID: L-2017-LLA-0307)
ML18012A0782018-01-12012 January 2018
NRR E-mail Capture - for Your Review - Draft RAI for Oyster Creek Permanently Defueled EP and EAL Scheme LAR (CAC No. MG0160; EPID: L-2017-LLA-0307)
ML18003A4122018-01-0303 January 2018
NRR E-mail Capture - for Your Action - RAI - Oyster Creek BWRVIP LAR (CAC No. MG0180; EPID No. L-LLA-2017-0304)
ML17320A6372017-11-16016 November 2017
NRR E-mail Capture - Oyster Creek Nuclear Generating Station - Request for Additional Information Regarding License Amendment Request for Emergency Plan Exemption (CAC No. MG0153; EPID L-2017-LLE-0020)
ML17319A9422017-11-15015 November 2017
NRR E-mail Capture - Oyster Creek Nuclear Generating Station - Request for Additional Information Regarding License Amendment Request for Emergency Plan Exemption (CAC No. MG0153; EPID L-2017-LLE-0020)
ML17208A7472017-09-14014 September 2017
Request for Additional Information, License Amendment Request to Revise the Site Emergency Plan for On-Shift and Emergency Response Organization Staffing for Permanently Defueled Condition
ML17136A3852017-05-12012 May 2017
NRR E-mail Capture - (External_Sender) Oyster Creek Response to NRC Flood MSA Audit Review Questions
ML17038A5152017-02-0303 February 2017
NRR E-mail Capture - (External_Sender) Oyster Creek Nuclear Generating Station Response to Environmental Questions Regarding the Mitigating Strategies Final Integrated Plan and SPF Instrumentation Compliance Letter
ML16244A3112016-10-0707 October 2016
Request for Additional Information Regarding License Amendment Request for Technical Specifications Section 6.0, Administrative Controls for Permanently Defueled Condition
ML16125A2452016-05-16016 May 2016
Request for Additional Information Regarding Request for Approval of a Certified Fuel Handler Training and Retraining Program
ML16060A1722016-03-0707 March 2016
Request for Additional Information Regarding Preliminary Decommissioning Cost Estimate and Spent Fuel Management Program
ML16056A0012016-02-24024 February 2016
NRR E-mail Capture - Request for Additional Information - Oyster Creek Security Plan - Revision 14
ML17144A3702015-05-0606 May 2015
NRC Email J. Hughey( NRC) to Dave Distel (Exelon)- Request for Additional Information Regarding Request for Extension to Comply with NRC Order EA-13-109, Dated May 6, 2015
ML15005A0692015-01-0606 January 2015
Request for Additional Information Regarding the Fourth 10-Year Interval Inservice Inspection Program Plan Requests for Relief
ML14246A5182014-09-17017 September 2014
Request for Additional Information Regarding the Fourth 10-Year Interval Inservice Inspection Program Plan Requests for Relief
ML14252A7032014-09-17017 September 2014
Request for Additional Information Regarding Preliminary Decommissioning Cost Estimate and Annual Decommissioning Funding Status Reports
ML14247A5862014-09-16016 September 2014
Request for Additional Information Regarding License Amendment Request to Revise Snubber Surveillance Requirements (Tac No. MF4124)
ML14114A5632014-09-10010 September 2014
Request for Additional Information Regarding License Amendment Request Related to Reactor Building Vital Area Access Control
ML14220A0272014-08-27027 August 2014
Request for Additional Information Regarding Request for Extension to Comply with NRC Order EA-13-109: Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions
ML14083A1602014-03-31031 March 2014
Request for Additional Information Regarding Bulletin 2012-01, Design Vulnerability in Electric Power System
ML13351A3142013-12-20020 December 2013
Request for Additional Information Regarding Response to Bulletin 2012-01, Design Vulnerability in Electric Power System.
ML13268A0312013-11-0808 November 2013
Interim Staff Evaluation and Request for Additional Information Regarding the Overall Integrated Plan for Implementation of Order EA-12-051, Reliable Spent Fuel Pool Instrumentation
ML13304B4182013-11-0101 November 2013
Request for Additional Information Associated with Near-Term Task Force Recommendation 2.3, Seismic Walkdowns
ML13227A3042013-08-28028 August 2013
Request for Additional Information Regarding Overall Integrated Plan for Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)
ML13196A3382013-07-15015 July 2013
2013 Decommissioning Fund Report RAI
ML12229A4332012-09-0606 September 2012
Request for Additional Information Regarding Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 12 (TAC No. D91660)
ML12056A0502012-03-12012 March 2012
Enclosure 4 - Recommendation 2.3: Flooding
ML12056A0482012-03-12012 March 2012
Enclosure 2 - Recommendation 2.1: Flooding
ML12056A0472012-03-12012 March 2012
Enclosure 1 - Recommendation 2.1: Seismic
ML12056A0492012-03-12012 March 2012
Enclosure 3 - Recommendation 2.3: Seismic
ML12056A0512012-03-12012 March 2012
Enclosure 5 - Recommendation 9.3: Emergency Preparedness
ML12032A1872012-02-21021 February 2012
Request for Additional Information Related to the Response to License Renewal Commitment No. 10 Regarding Thermal Aging Irradiation Embrittlement of Cast Austenitic Stainless Steel
ML12025A2542012-01-23023 January 2012
Request for Additional Information - License Renewal Commitment Regarding Thermal Aging and Neutron Irradiation Embrittlement of CASS Materials
ML1200603242012-01-19019 January 2012
Fleet, RAI, Proposed Alternative to Use American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Case N-789 (N-789)
ML1131200572011-11-22022 November 2011
Fleet - Bulletin 2011 01, Mitigating Strategies Request for Additional Information Regarding 60-Day Response
ML1123604472011-08-24024 August 2011
E-mail, Miller to Loomis, Oyster Creek Flaw Evaluation Draft RAI Questions
ML1120910012011-07-22022 July 2011
Generation Station - Licensed Operator Request for Additional Information
ML1116805642011-07-0707 July 2011
Fleet, Request for Additional Information, 2011 Report on Status of Decommissioning Funding for Reactors
2021-04-08
[Table view] |
Text
August 11,2006 LICENSEE: AmerGen Energy Company, LLC FACILITY: Oyster Creek Nuclear Generating Station
SUBJECT:
SUMMARY
OF A TELEPHONE CONFERENCE CALL HELD ON MARCH 16, 2006, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND AMERGEN ENERGY COMPANY, LLC, CONCERNING DRAFT REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL APPLICATION The U.S. Nuclear Regulatory Commission staff (NRC or the staff), and representatives of AmerGen Energy Company, LLC (AmerGen), held a telephone conference call on March 16, 2006, to discuss and clarify the staffs 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 staffs D-RAIs. provides a listing of the conference call participants. Enclosure 2 contains a listing of 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 Manager License Renewal Branch A Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosures:
As stated cc w/encls: See next page
ML062230429 DOCUMENT NAME: E:\Filenet\ML062230429.wpd OFFICE PM:RLRA:DLR LA:RLRA:DLR BC:RLRA:DLR NAME DAshley YEdmonds LLund DATE 08/ 11 /06 08/ 10 /06 08/ 11 /06 cc:
Site Vice President - Oyster Creek Director - Licensing and Regulatory Affairs Nuclear Generating Station AmerGen Energy Company, LLC AmerGen Energy Company, LLC Correspondence Control P.O. Box 388 P.O. Box 160 Forked River, NJ 08731 Kennett Square, PA 19348 Senior Vice President of Manager Licensing - Oyster Creek Operations Exelon Generation Company, LLC AmerGen Energy Company, LLC Correspondence Control 200 Exelon Way, KSA 3-N P.O. Box 160 Kennett Square, PA 19348 Kennett Square, PA 19348 Kathryn M. Sutton, Esquire Regulatory Assurance Manager Morgan, Lewis, & Bockius LLP Oyster Creek 1111 Pennsylvania Avenue, NW AmerGen Energy Company, LLC Washington, DC 20004 P.O. Box 388 Forked River, NJ 08731 Kent Tosch, Chief New Jersey Department of Assistant General Counsel Environmental Protection AmerGen Energy Company, LLC Bureau of Nuclear Engineering 200 Exelon Way CN 415 Kennett Square, PA 19348 Trenton, NJ 08625 Ron Bellamy, Region I Vice President - Licensing and U.S. Nuclear Regulatory Commission Regulatory Affairs 475 Allendale Road AmerGen Energy Company, LLC King of Prussia, PA 19406-1415 4300 Winfield Road Warrenville, IL 60555 Correspondence Control Desk AmerGen Energy Company, LLC Regional Administrator, Region I 200 Exelon Way, KSA 1--1 U.S. Nuclear Regulatory Commission Kennett Square, PA 19348 475 Allendale Road King of Prussia, PA 19406-1415 Oyster Creek Nuclear Generating Station Plant Manager Mayor of Lacey Township AmerGen Energy Company, LLC 818 West Lacey Road P.O. Box 388 Forked River, NJ 08731 Forked River, NJ 08731 Senior Resident Inspector License Renewal Manager U.S. Nuclear Regulatory Commission Exelon Generation Company, LLC P.O. Box 445 200 Exelon Way, Suite 230 Forked River, NJ 08731 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
Note to: AmerGen Energy Company, LLC, Facility: Oyster Creek Nuclear Generating Station from Donnie Ashley dated August 11, 2006.
SUBJECT:
SUMMARY
OF A TELEPHONE CONFERENCE CALL HELD ON MARCH 16, 2006, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND AMERGEN ENERGY COMPANY, LLC, CONCERNING DRAFT REQUEST FOR ADDITIONAL INFORMATION PERTAINING TO THE OYSTER CREEK NUCLEAR GENERATING STATION, LICENSE RENEWAL APPLICATION DISTRIBUTION:
HARD COPY DLR R/F E-MAIL:
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 DAshley VRodriguez RLaufer GMiller RBellamy, RI RCureton, RI JLilliendahl, RI MModes, RI MSykes, RI AHodgdon RidsOpaMail
LIST OF PARTICIPANTS FOR TELEPHONE CONFERENCE CALL TO DISCUSS THE OYSTER CREEK NUCLEAR GENERATING STATION LICENSE RENEWAL APPLICATION March 16, 2006 Participants Affiliations Donnie Ashley U.S. Nuclear Regulatory Commission (NRC)
Matt Mitchell NRC Chris Sydnor NRC Ganesh Cheruvenki NRC George Beck AmerGen Energy Company, LLC (AmerGen)
Mike May AmerGen Gary Stevens AmerGen Greg Harttraft AmerGen ENCLOSURE 1
DRAFT REQUESTS FOR ADDITIONAL INFORMATION (D-RAIs)
OYSTER CREEK NUCLEAR GENERATING STATION LICENSE RENEWAL APPLICATION March 16, 2006 The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of AmerGen Energy Company, LLC (AmerGen), held a telephone conference call on March 16, 2006, to discuss and clarify the staffs 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 4.2.2-1 Please 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 a formal RAI.
D-RAI 4.2.2-2 In 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 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.
Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI.
ENCLOSURE 2
D-RAI 4.2.7-1 Boiling Water Reactor Vessel and Internals Project (BWRVIP)-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.
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 1020 n/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 drytubes, or top guide.
Please clarify why there is no TLAA for the core shroud, incore instrumentation drytubes, 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.
Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI.
D-RAI 4.7.4-1 In Section 4.7.4, of the LRA Reactor Vessel Weld Flaw Evaluations, 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.
Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI.
D-RAI 4.7.5-1 In Section 4.7.5, of the LRA Control Rod Drive (CRD) Stub Tube Flaw Analysis, includes a discussion of several cracks that were found in the CRD stub tubes during construction and a subsequent repair of the cracks that included grinding out 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.
Discussion: The applicant indicated that the question is clear. This D-RAI will be sent as a formal RAI.