Letter Sequence Response to RAI |
---|
|
Initiation
- Request, Request, Request, Request
- Acceptance, Acceptance
- Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement
|
MONTHYEARML0720002812007-07-13013 July 2007 License Amendment Request - Stretch Power Uprate - Supplemental Information Project stage: Supplement ML0720003862007-07-13013 July 2007 License Amendment Request, Stretch Power Uprate Project stage: Request ML0726702162007-10-15015 October 2007 Stretch Power Uprate Acceptance Review Project stage: Acceptance Review ML0727005452007-10-18018 October 2007 Westinghouse Electric Company LLC Request for Withholding Information from Public Disclosure Regarding Letter (CAW-07-2296), Subject: NEU-07-128 P-Attachment, Millstone Nuclear Power Station Unit 2 Stretch Power Uprate Program Supplemental Project stage: Withholding Request Acceptance ML0729601792007-10-29029 October 2007 Request for Additional Information, Stretch Power Uprate Amendment Request Project stage: RAI ML0732309762007-11-19019 November 2007 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Project stage: Response to RAI ML0731706652007-11-26026 November 2007 Request for Additional Information on Instrumentation and Controls and Fire Protection Regarding Stretch Power Uprate Amendment Request Project stage: RAI ML0734802402007-12-13013 December 2007 Supplemental Information for License Amendment Request Stretch Power Uprate Project stage: Supplement ML0733703842007-12-14014 December 2007 Request for Additional Information on the Stretch Power Uprate Amendment Request Project stage: RAI ML0732500812007-12-14014 December 2007 Westinghouse Electric Company LLC Request for Withholding Information from Public Disclosure Regarding Letter (CAW-07-2252), Subject: WCAP-16271-P, Millstone Unit 3 Spent Fuel Pool Criticality Safety Analysis, March 2007 (Proprietary), Date Project stage: Withholding Request Acceptance ML0735200512007-12-17017 December 2007 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Project stage: Response to RAI ML0801006002008-01-10010 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions EEEB-07-0049 Through EEEB-07-0057 Project stage: Response to RAI ML0801006112008-01-10010 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question CPNB-07-0048 Project stage: Response to RAI ML0801006042008-01-10010 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question AADB-07-0012 Project stage: Response to RAI ML0801006062008-01-10010 January 2008 Response to Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions SCVB-07-0058 and SCVB-07-0059 Project stage: Response to RAI ML0805804762008-01-11011 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions SRXB-07-0013 Through SRXB-07-0047 Project stage: Response to RAI ML0801704952008-01-11011 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question EMCB-07-0070 Project stage: Response to RAI ML0801400772008-01-11011 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request, Response to Questions SBPB-07-0082 Through SBPB-07-0087 Project stage: Response to RAI ML0801106952008-01-11011 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request, Response to Questions CSGB-07-0010 and CSGB-07-0011 Project stage: Response to RAI ML0801405702008-01-14014 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions EMCB-07-0060 Through EMCB-07-0069 and EMCB-07-0071 Through EMCB-07-0081 Project stage: Response to RAI ML0802205062008-01-18018 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question EICB-07-0106 Project stage: Response to RAI ML0802205272008-01-18018 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question SBPB-07-0105 Project stage: Response to RAI ML0802205302008-01-18018 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions SCVB-07-0091 Through SCVB-07-0104 Project stage: Response to RAI ML0802205412008-01-18018 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Questions SRXB-07-0088 Through SRXB-07-0090 Project stage: Response to RAI ML0802803752008-01-18018 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request, Response to Question AADB-07-0107 Project stage: Response to RAI ML0803203082008-01-31031 January 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question EMCB-07-0070 Project stage: Response to RAI ML0805603922008-02-25025 February 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request, Supplemental Response to Question EMCB-07-0072 Project stage: Supplement ML0805606152008-02-25025 February 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Revised and Supplemental Responses to Questions AFPB-07-0007 & AFPB-07-0008 Project stage: Supplement ML0807103912008-03-10010 March 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question SCVB-07-0091 Project stage: Response to RAI ML0807103772008-03-10010 March 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Response to Question AADB-07-0107 Project stage: Response to RAI ML0803501322008-03-13013 March 2008 Request for Withholding Information from Public Disclosure Regarding Millstone Power Station, Unit 3 Project stage: Withholding Request Acceptance ML0808405272008-03-17017 March 2008 Nancy Burton Petition to Intervene and Request for Hearing Project stage: Request ML0808508942008-03-25025 March 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Supplemental Response to Question EEEB-07-0052 Project stage: Supplement ML0808802682008-03-27027 March 2008 Supplemental Information Regarding Stretch Power Uprate License Amendment Request Project stage: Supplement ML0804606202008-03-28028 March 2008 Potential Schedule Change for Proposed Stretch Power Uprate License Amendment Request Project stage: Other ML0814300142008-04-0404 April 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Supplemental Response to Question EMCB-07-0072 Project stage: Supplement ML0810800242008-04-23023 April 2008 Request for Additional Information, Stretch Power Uprate Amendment Request Project stage: RAI ML0811506792008-04-24024 April 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Supplemental Information, Rod Withdrawal at Power Event Project stage: Supplement ML0812006432008-04-29029 April 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Revised Response to Questions EEEB-07-0052, EEEB-07-0054 and EEEB-07-0055 Project stage: Response to RAI ML0813606252008-05-15015 May 2008 Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Revised Response to Follow-ups EEEB-08-0108 Through EEEB08-0113 to Question EEEB-07-0052 Project stage: Response to RAI ML0814204432008-05-20020 May 2008 Supplemental Information Regarding Stretch Power Uprate License Amendment Request Miscellaneous Updates to the License Amendment Request Project stage: Supplement ML0814208242008-05-21021 May 2008 Stretch Power Uprate License Amendment Request Additional Information in Connection with the NRC Audit Held on May 13, 2008 in Rockville, Maryland Project stage: Request ML0808704572008-05-29029 May 2008 Draft Environmental Assessment and Finding of No Significant Impact for Proposed Stretch Power Uprate Project stage: Draft Other ML0817502222008-06-16016 June 2008 Notice of Appeal Project stage: Request ML0819302742008-07-10010 July 2008 Stretch Power Uprate License Amendment Request DNC Comments on Draft Safety Evaluation - Stretch Power Uprate Project stage: Draft Request ML0819307572008-07-11011 July 2008 Follow-up to Dominion Nuclear Connecticut, Inc. Letter, Dated July 10, 2008, Regarding the Draft Safety Evaluation - Stretch Power Uprate Project stage: Draft Approval ML0819901122008-07-16016 July 2008 Stretch Power Uprate License Amendment Request - Supplement to DNC Comments on Draft Safety Evaluation - Stretch Power Uprate ML0819300702008-07-30030 July 2008 Final Environmental Assessment and Finding of No Significant Impact for Proposed Stretch Power Uprate Project stage: Other ML0816105852008-08-12012 August 2008 License Amendment, Stretch Power Uprate Project stage: Other ML0822504182008-08-12012 August 2008 Stretch Power Uprate, Tec Specs to Amd 242 Project stage: Acceptance Review 2008-03-17
[Table View] |
|
---|
Category:Letter
MONTHYEARIR 05000336/20240032024-11-0707 November 2024 Integrated Inspection Report 05000336/2024003 and 05000423/2024003 and Apparent Violation and Independent Spent Fuel Storage Installation Inspection Report 07200008/2024001 ML24289A0152024-10-21021 October 2024 Review of the Fall 2023 Steam Generator Tube Inspection Report 05000423/LER-2024-001, Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary2024-10-14014 October 2024 Loss of Safety Function and Condition Prohibited by Technical Specifications for Loss of Secondary Containment Boundary IR 05000336/20244022024-10-0808 October 2024 Security Baseline Inspection Report 05000336/2024402 and 05000423/2024402 (Cover Letter Only) ML24281A1102024-10-0707 October 2024 Requalification Program Inspection 05000423/LER-2023-006-02, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-09-26026 September 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24240A1692024-09-18018 September 2024 Cy 2023 Summary of Decommissioning Trust Fund Status ML24260A2192024-09-16016 September 2024 Decommissioning Trust Fund Disbursement - Revision to Previous Thirty-Day Written Notification ML24260A1952024-09-16016 September 2024 Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome Gaia Fuel ML24248A2272024-09-0404 September 2024 Operator Licensing Examination Approval ML24240A1532024-09-0303 September 2024 Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome Gaia Fuel IR 05000336/20240052024-08-29029 August 2024 Updated Inspection Plan for Millstone Power Station, Units 2 and 3 (Reports 05000336/2024005 and 05000423/2024005 IR 05000336/20240022024-08-13013 August 2024 Integrated Inspection Report 05000336/2024002 and 05000423/2024002 ML24221A2872024-08-0808 August 2024 Independent Spent Fuel Storage Installation (ISFSI) - Submittal of Cask Registration for Spent Fuel Storage IR 05000336/20244412024-08-0606 August 2024 Supplemental Inspection Report 05000336/2024441 and 05000423/2024441 and Follow-Up Assessment Letter (Cover Letter Only) ML24212A0742024-08-0505 August 2024 Request for Withholding Information from Public Disclosure - Millstone Power Station, Unit No. 3, Proposed Alternative Request IR-4-13 to Support Steam Generator Channel Head Drain Modification ML24211A1712024-07-25025 July 2024 Associated Independent Spent Fuels Storage Installation, Revision to Emergency Plan - Report of Change IR 05000336/20244032024-07-22022 July 2024 Information Request for the Cybersecurity Baseline Inspection, Notification to Perform Inspection 05000336/2024403 and 05000423/2024403 IR 05000336/20245012024-07-0101 July 2024 Emergency Preparedness Biennial Exercise Inspection Report 05000336/2024501 and 05000423/2024501 ML24180A0932024-06-28028 June 2024 Readiness for Additional Inspection: EA-23-144 IR 05000336/20240102024-06-26026 June 2024 Biennial Problem Identification and Resolution Inspection Report 05000336/2024010 and 05000423/2024010 ML24178A2422024-06-25025 June 2024 2023 Annual Report of Emergency Core Cooling System (ECCS) Model, Changes Pursuant to the Requirements of 10 CFR 50.46 IR 05000336/20244402024-06-24024 June 2024 Final Significance Determination for Security-Related Greater than Green Finding(S) with Assessment Follow-up; IR 05000336/2024440 and 05000423/2024440 and Notice of Violation(S), NRC Investigation Rpt 1-2024-001 (Cvr Ltr Only) ML24176A1782024-06-20020 June 2024 Update to the Final Safety Analysis Report ML24176A2622024-06-20020 June 2024 Update to the Final Safety Analysis Report, Revision 37 ML24177A2792024-06-20020 June 2024 Preparation and Scheduling of Operator Licensing Examinations ML24280A0012024-06-20020 June 2024 Update to the Final Safety Analysis Report (Redacted Version) ML24281A2072024-06-20020 June 2024 Update to the Final Safety Analysis Report, Revision 37 (Redacted Version) 05000336/LER-2024-001, Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications2024-06-10010 June 2024 Control Room Air Conditioning Unit Inoperable Due to Refrigerant Overcharge Resulting in a Condition Prohibited by Technical Specifications ML24170B0532024-06-10010 June 2024 DOM-NAF-2-P/NP-A, Revision 0.5, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code ML24165A1292024-06-0505 June 2024 ISFSI, 10 CFR 50.59 Annual Change Report for 2023 Annual Regulatory Commitment Change Report for 2023 ML24128A2772024-06-0404 June 2024 Issuance of Amendment No. 290 to Revise TSs for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report for Use of Framatome Gaia Fuel (EPID L-2023-LLA-0074) (Non-Proprietary) ML24151A6482024-06-0303 June 2024 Changes in Reactor Decommissioning Branch Project Management Assignments for Some Decommissioning Facilities ML24110A0562024-05-21021 May 2024 Exemption from the Requirements of 10 CFR Part 50, Section 50.46, and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013) (Letter) ML24109A0032024-05-21021 May 2024 Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary) ML24142A0952024-05-20020 May 2024 End of Cycle 22 Steam Generator Tube Inspection Report 05000423/LER-2023-006-01, Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-20020 May 2024 Pressurizer Power Operated Relief Valve Failed to Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications ML24141A2432024-05-20020 May 2024 Response to Request for Additional Information Regarding Alloy 600 Aging Management Program Submittal Related to License Renewal Commitment No. 15 IR 05000336/20240012024-05-14014 May 2024 Integrated Inspection Report 05000336/2024001 and 05000423/2024001 and Apparent Violation ML24123A2042024-05-0202 May 2024 Pre-Decisional Replay to EA-23-144 05000423/LER-2023-006, Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications2024-05-0202 May 2024 Pressurizer Power Operated Relief Valve Failed to Stroke Open During Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications IR 05000336/20244012024-04-30030 April 2024 Security Baseline Inspection Report 05000336/2024401 and 05000423/2024401 (Cover Letter Only) ML24123A1222024-04-30030 April 2024 Inservice Inspection Program - Owners Activity Report, Refueling Outage 22 ML24116A0452024-04-25025 April 2024 Special Inspection Follow-Up Report 05000336/2024440 and 05000423/2024440 and Preliminary Finding(S) of Greater than Very Low Significance and NRC Investigation Report No. 1-2024-001 (Cover Letter Only) ML24116A1742024-04-24024 April 2024 Annual Radiological Environmental Operating Report ML24114A2662024-04-24024 April 2024 Submittal of 2023 Annual Radioactive Effluent Release Report ML24103A0202024-04-22022 April 2024 Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits ML24106A2032024-04-15015 April 2024 2023 Annual Environmental Operating Report ML24088A3302024-04-0404 April 2024 Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome Gaia Fuel ML24093A2162024-04-0101 April 2024 Response to Request for Additional Information Regarding License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits 2024-09-04
[Table view] |
Text
Dominion Nuclear Connecticut, Inc.';000 Dominion Boulevard, Glen Allen, Virginia 21060 Wch Address: www.dom.com November 19, 2007 07-0751 RO 50-423 NPF-49 Serial No.: NLOS/MAE: Docket No.: License No.: U.S.Nuclear Regulatory Commission Attention:
Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2378 DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST Dominion Nuclear Connecticut, Inc.(DNC)submitted a stretch power uprate license amendment request for Millstone Power Station Unit 3 (MPS3)in letters dated July 13, 2007 (Serial Nos.07-0450 and 07-0450A)and supplemented the submittal by letter dated September 12, 2007 (07-04508).
The NRC staff forwarded a request for additional information (RAI)in an October 29, 2007 letter.The response to this RAI is provided in the attachment to this letter.The information provided by this letter does not affect the conclusions of the significant hazards consideration discussion in the July 13, 2007 ONe letter (Serial No.07-0450).Should you have any questions in regard to this submittal, please contact Ms.Margaret Earle at 804-273-2768.
Very truly yours, Vice President Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T.Bischof, who is Vice President Nuclear Engineering of Dominion Nuclear Connecticut, Inc.He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.Acknowledged before me this/q"h, day of,2007.My Commission Expires:
3/,.;"008.MARGARET*.BENNE" Notary PublIC Commonwealth of VIrginia I My Commluton ExpIfeI Aug S 1, 2001-------
Commitments made in this letter: None Attachment cc: U.S.Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr.J.G.Lamb U.S.Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-881A Rockville, MD 20852-2738 Ms.C.J.Sanders Project Manager U.S.Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-883 Rockville, MD 20852-2738 Mr.S.W.Shaffer NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Serial No.07-0751 Docket No.50-423 RAI, Stretch Power Uprate bc page 2 of 2 Serial No.07-0751 Docket No.50-423 Attachment, page 1 of 6 ATTACHMENT LICENSE AMENDMENT REQUEST STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No.07-0751 Docket No.50-423 Attachment, page 2 of 6 Stretch Power Uprate License Amendment Request Response to Request for Additional Information Dominion Nuclear Connecticut, Inc.(DNC)submitted a stretch power uprate (SPU)license amendment request for Millstone Power Station Unit 3 (MPS3)in letters dated July 13, 2007 (Serial Nos.07-0450 and 07-0450A)and supplemented the submittal by letter dated September 12, 2007 (Serial No.0450S).The NRC staff forwarded a request for additional information (RAI)in an October 29,2007 letter.The response to the RAI is provided below.NRC Question CVIB-07-001 Title 10 of the Code Federal Code of Federal Regulations (10 CFR)Part 50, Appendix G, provides fracture toughness requirements for ferritic materials (low alloy steel or carbon steel)materials in the reactor coolant pressure boundary (RCPS)components.
10 CFR Part 50, Appendix G also identifies that RCPS materials must satisfy the criteria in Appendix G of Section XI of the American Society of Mechanical Engineers Soiler and Pressure Vessel Code (ASME Code)to ensure the structural integrity of the ferritic components of the RCPS during any comdition of normal operation, including anticipated operational occurrences and hydrostatic tests.Consistent with the requirements specified in 10 CFR Part 50, Appendix G, the licensee performed fracture toughness analyses of the ferritic reactor vessel (RV)materials.
The staff requests that the licensee confirm that these analyses bound the fracture toughness requirements of the ferritic RCPS components other than the RV components at MPS, Unit 3 ONe Response For the ferritic RCPS components other than the reactor vessel (Le., pressurizer and steam generators), the fracture toughness requirements were addressed in the original ASME Code analyses.The results of these Code required 10 CFR 50 Appendix G analyses are documented in each component's Code stress report.The changes associated with SPU have no impact on the 10 CFR 50 Appendix G analyses documented in the pressurizer and steam generator stress reports.Therefore, the fracture toughness requirements associated with 10CFR50 Appendix G for the ferritic RCPS components other than the reactor vessel will continue to be satisfied after implementation of SPU.
Serial No.07-0751 Docket No.50-423 Attachment, page 3 of 6 NRC Question CVIB-07-002 Section 2.1.2.2-1 delineates the difference between 1/4T neutron fluence (E>1 MeV)estimated in the license extension application and the value proposed in the current application for SPU, i.e., 1.97x10 19 n/cm 2 and 1.63x10 19 n/cm 2 , respectively.
Similarly, Section 2.1.3.2.3 indicates a similar difference between the peak inside surface vessel fluence to 54 effective full power years (EFPYs)of operation.
Apparently, both calculations were carried out using methodologies adhering to the guidance in RG 1.190,"Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Please describe the physical reasons that justify the lower fluence value for the 54 EFPYs under SPU conditions.
ONe Response The change in estimated neutron fluence (both surface and 1/4T)is due to new information received from analysis of surveillance capsule W removed in October 2005 at 13.8 EFPYs, and from recent fluence analyses performed in 2006 to support the SPU.The neutron fluence estimates cited in the license renewal application were based on projections from surveillance capsule X removed at 8.0 EFPYs.The recent fluence calculations performed for capsule Wand SPU were based on three-dimensional synthesized neutron flux distributions, where forward transport calculations were performed in (r,8), (r,z), and (r)geometry.The fluence calculations performed for capsule X were based on two-dimensional (r,8)analyses using forward transport and adjoint calculations.
As noted in Section 1.3.4 of Regulatory Guide 1.190, the use of a synthesized three-dimensional approach, with both radial and axial transport calculations, is generally more accurate.Additional details of the neutron fluence analyses are provided below.The reactor vessel inner surface neutron fluence (E>1 MeV)of 1.97x10 19 n/cm 2 at 32 EFPY, as cited in the license renewal application, is discussed in15405, Revision 0"Analysis of Capsule X from Northeast Nuclear Energy Company Millstone Unit 3 Reactor Vessel Radiation Surveillance Program," May 2000.WCAP-15405 was submitted to the NRC on May 17, 2000,"Millstone Nuclear Power Station Unit No.3 Submittal of Second Reactor Vessel Surveillance Capsule Report" (ADAMS Accession No.ML003717395).
This WCAP documents the analysis and evaluation of surveillance capsule X removed at the end of the sixth cycle after 8.0 EFPY.The results of the evaluation of capsule X were used to project the estimated fluence from cycle 6 to the end of life at 32 EFPY (40-year operation), and at 54 EFPY for 60-year operation.
Very conservative assumptions were made for estimating the end of life fluence.Projections for future operation were based on the assumption that Serial No.07-0751 Docket No.50-423 Attachment, page 4 of 6 neutron flux values from Cycles 4 through 6 would continue to be applicable throughout plant life.From Table 6-2 of WCAP-15405, the average neutron flux for Cycles4-6 is 1.93x10 10 n/cm 2-sec at the limiting azimuthal location.This equilibrium flux value was used in the fluence projections.
From Table 6-13 of WCAP-15405 the highest calculated inner surface fluence at 32 EFPY was 1.97x10 19 n/cm 2.At 54 EFPY, the calculated fluence was 3.31x10 19 n/cm 2.In October 2005, at the end of cycle 10, surveillance capsule W was removed for evaluation and analysis.The results of the evaluations and analysis of this surveillance capsule are summarized in WCAP-16629-NP,"Analysis of Capsule W from the Dominion Nuclear Operating Company Millstone Unit 3 Reactor Vessel Radiation Surveillance Nuclear Program," Revision 0, September 2006.WCAP-16629-NP was submitted to the NRC on October 2, 2006,"Dominion Nuclear Connecticut, Inc.Millstone Power Station Unit 3 Submittal of Third Reactor Surveillance Capsule Report" (ADAMS Accession No.ML062850221).
In addition to the evaluation and analysis of the surveillance capsule, this WCAP also contains projections of the maximum fluence from Cycle 10 to end of life.From Table 6-2 of WCAP-16629-NP, the maximum fluence was estimated to be 1.75x10 19 n/cm 2 at 32 EFPY, and 2.96x10 19 n/cm 2 at 54 EFPY.These projections were carried out based on an assumed core power uprate from 3411 MWt to 3650 MWt at the onset of Cycle 11.The projections were based on the assumption that the relative spatial core power distributions and associated plant operating characteristics from Cycle 10 were representative of future plant operation.
For Cycles 11 and beyond, the equilibrium flux values are based on Cycle 10 plus a multiplier of 1.07 to account for the assumed power uprate.It is worth noting that the equilibrium flux values shown in Table 6-2 of NP arelessthan those reported in WCAP-15405.
This can be attributed to the continued used of low leakage core designs at Millstone Unit 3.Additionally, added refinements in the fluence calculation methodology were used in16629-NP as compared to the methodology used in WCAP-15405.
The fluence analysis per WCAP-16629-NP employedaP s legendre expansion and an S16 angUlar quadrature discretization, whereas the fluence analyses per15405 employedaP 3 legendre expansion and an Sa angular quadrature discretization.
For the Stretch Power Uprate (SPU)analysis, these neutron fluence projections have been updated to reflect the current uprate schedule.Preliminary designs of transition and equilibrium fuel cycles for the SPU program have been completed.
These designs describe a pre-uprate Cycle 12 operating at 3411 MWt, two transition cycles (Cycles 13 and 14)operating at 3650 MWt (SPU power level)and an equilibrium fuel cycle (Cycle 15)also operating at 3650 MWt.Using these designs the projected maximum inner surface fluence at 32 EFPY is 1.63x1 0 19 n/cm 2 , and at 54 EFPY is 2.70x1 0 19 n/cm 2.
Serial No.07-0751 Docket No.50-423 Attachment, page 5 of 6 The neutron transport methodology used in the Capsule Wand SPU calculations follows the guidance and meets the requirements of Regulatory Guide 1.190"Calculational and Dosimetry Methods for Determining Pressure Vessel NeutronFluence."The Capsule X fluence analyses per WCAP-15405 were compliant with the requirements of Draft Regulatory Guide DG-1053, which was the precursor to Regulatory Guide 1.190.Additionally, the methods used to determine the pressure vessel neutron exposure make use of the NRC approved methodology described in WCAP-14040-NP-A"Methodology Used to Develop Cold Overpressurization Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.NRC Question CVIB-07-003 In 2001, one Babcock and Wilcox (B&W)licensee experienced failures of two control rod drive mechanism 17-4 precipitation hardened (PH)lead screw male couplings.
The failures were attributed to thermal embrittlement of the 17-4 PH martensitic stainless steel materials.
The staff issued NRC Information Notice (IN)2007-02,"Failure of Control Rod Drive Mechanism Lead Screw Male Coupling at a Babcock and Wilcox Designed Facility." In IN 2007-02, the staff reiterated the importance of implementing frequent visual andsurfaceexaminations for identifying thermal embrittlement in 17-4 PH martensitic stainless steel reactor vessel internals (RVI)components.
The staff requests that the licensee provide the following information with respect to monitoring the aging degradation of the 17-4 PH martensitic stainless steel materials used in RVI components at MPS3: (1)Identify 17-4 PH martensitic stainless steel RVI components at MPS3, if any.(2)Identify the method of inspection that was performed thus far on these components and provide information regarding any aging degradation that was identified thus far in these components.
ONC Response (1)There are no 17-4 PH materials used in the reactor vessel internals.
(2)This question is not applicable since there are no 17-4 PH materials used in the reactor vessel internals.
Serial No.07-0751 Docket No.50-423 Attachment, page 6 of 6 NRC Question SSPS-07-004 In Attachment 5, on page 2.5-114, under the section"Turbine Driven Feedwater Pump Turbine Control Valves," the licensee stated in the application that"[the]engineering evaluation to confirm whether or not more steam flow is required for turbine driven feedwater pump turbines for SPU conditions is in progress." The NRC staff cannotinitiatea review of this section until the results of the evaluation and a description of any impacts or modifications are submitted.
Please provide the results of the evaluation and a description of any impacts or modifications.
ONC Response The GE engineering evaluations of the need for thefeedwaterturbine control valve modifications have been completed.
The evaluations concluded that no hardware modifications to those valves are necessary in support of the SPU.NRC Question SSPS-07-005 In Attachment 5, on page 2.5-22, under the section,Turbine Results," DNC performed high-pressure turbine rotor phased array testing of the tangential-entry dovetail regions of the wheel rim, which revealed indications on the turbine first stage wheel.DNC is currently evaluating either installing a new high-pressure rotor, or disassembling first stage buckets, conducting inspections, removing indications and possible repairs.ONC needs to identify their proposed resolution and affects on the proposed SPU before the NRC staff can complete its review.ONC Response During the next refueling outage (October 2008)DNC will be removing buckets from both of the first stage wheels of the Millstone HP Rotor.The wheels will be machined to a new bucket attachment dovetail configuration.
This machining will remove the subject wheel indications.
New longer shank buckets will be installed on the machined wheel.The new dovetail configuration and buckets are being designed by the turbine OEM for the SPU conditions.
This repair will reestablish the normal 10-year HP Rotor inspection interval.