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MONTHYEARML0615702092006-05-26026 May 2006 Additional Information to Support the Request to Extend the Second 10-Year, American Society of Mechanical Engineers Section IX, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations - Relief Request No. 34 Project stage: Request ML0624905132006-09-20020 September 2006 Relief, Request to Extend the Second 10-Year Inservice Inspection Program Interval Project stage: Other 2006-05-26
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Category:Code Relief or Alternative
MONTHYEARML22124A2412022-05-12012 May 2022 Relief Request 67 for an Alternate Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection ML22049A0572022-02-23023 February 2022 Relief Request 69 to Extend Inservice Inspection of Containment Tendon by Four Months Due to the Covid 19 Pandemic (EPID L 2022 Llr 0011 (Covid 19)) ML21278B0912021-10-0606 October 2021 Presubmittal Meeting with Arizona Public Service Company Regarding Relief Request No. 68 for Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (EPID L-2021-LRM-0104) (Slides) ML21228A1042021-08-12012 August 2021 RR-67, Request for Alternative Frequency to Containment Unbonded Post-Tensioning System Inservice Inspection (EPID L-2021-LLR-0050) (Email) ML21089A0102021-04-0202 April 2021 Relief Request 66 to Defer Inservice Inspection of Containment Tendon by One Year Due to Covid-19 Pandemic ML20088A5332020-03-27027 March 2020 Relief Request 65 - Unit 2, COVID-19, Request for Relief from Bottom Mounted Instrumentation Nozzles and a Pressurizer Nozzle to Surge Line Weld Overlay Examination ML19263F8752019-09-20020 September 2019 Relief Request VRR-01: Request for Alternative Frequency to Supplemental Indication Requirements of 10 CFR 50.55a(b)C3)(xi) ML19107A3722019-04-19019 April 2019 Relief Request 62 Regarding Proposed Alternative Pressurizer Heater Sleeve Repairs ML19044A6412019-02-19019 February 2019 Relief Request No. 58 for the Third 10-Year Inservice Inspection Interval, Request for Relief from the American Society of Mechanical Engineers for Certain Class 1 and Class 2 Welds ML18285A0292018-10-19019 October 2018 Relief Request 59 for the Deferral of Reactor Vessel Beltline Region Interior Attachment Examinations ML18079A7252018-03-17017 March 2018 Relief Request 58 - Unit 2 Impractical Examinations for the Third 10-Year Inservice Inspection Interval ML18016A1722018-01-12012 January 2018 Relief Request 50 - Request for Alternative to American Society of Mechanical Engineers Section XI Requirements for Pressure Retaining Boundary During System Leakage Tests ML17074A2092017-03-16016 March 2017 Relief Request GRR-01 to ASME Code Case OMN-20 for Third 10-Year Interval Pump and Valve Inservice Testing Program ML16172A0382016-06-23023 June 2016 Relief Request 54 to Approve an Alternative to Flaw Removal for Reactor Coolant Pump 2A Suction Pressure Instrument Nozzle, for the Third 10-Year Inservice Inspection Interval ML15238B6612015-09-15015 September 2015 Relief Request 53, Alternative to ASME Code Section XI Requirements Related to Flaw Removal ML15079A0062015-03-30030 March 2015 Relief Request 52 - Request for Approval of Alternate to Flaw Removal, Flaw Characterization and Successive Examinations, for the Remainder of Its Useful Life ML14093A4072014-04-10010 April 2014 Relief Request 51, Alternative to ASME Code for Flaw Removal, Reactor Vessel Bottom-mounted Instrumentation Nozzle, Third 10-Year ISI Interval ML13091A1772013-04-12012 April 2013 Relief Request 48 - Alternative to ASME Section III, Phased Array Ultrasonic Exam Techniques in Lieu of Radiography for Remainder of Third 10-Year Inservice Inspection Interval ML13085A2542013-04-0404 April 2013 Relief Request 49, Alternative to ASME Code, Section XI, for Reactor Vessel Head Flange Seal Leak Detection Piping, Third 10-Year Inservice Inspection Interval ML12257A1412012-09-18018 September 2012 Request for Relief to Use Subsequent Edition and Addenda of ASME Code, Section XI, for Examination Categories B-L-1, B-M-1, and C-G, for the Third 10-Year Inservice Inspection Interval ML1026700802010-10-14014 October 2010 Relief Request No. 47, from Certain ASME Code Requirements for Reactor Vessel Nozzle to Vessel Welds for Second 10-Year Inservice Inspection Interval ML1025001432010-09-21021 September 2010 Relief Request No. 45 from Certain ASME Code Class 1 Weld and Component Volumetric Exams of Reactor Vessel Outlet Nozzles for Second 10-Year Inservice Inspection Interval ML1021604872010-08-27027 August 2010 Relief Request No. RR-44, Reactor Vessel Weld Visual Examination Interval Extension for Third 10-Year Inservice Inspection Interval ML0921603982009-07-17017 July 2009 Submittal of Relief Request from the American Society of Mechanical Engineers (ASME) Code, Section XI - Relief Request No. 46 ML0917001972009-07-0202 July 2009 Relief Request No. 39, Alternative to Repair Weld Methods, ASME Code, Section XI for Remainder of Operating License ML0916304492009-06-0202 June 2009 Inservice Testing Relief Request for High Pressure Safety Injection Pump Testing - Pump Relief Request PRR-08, Revision 1 ML0905801992009-02-17017 February 2009 Relief Request 41, to Use Appendix I of ASME Code Case N-729-1 ML0830105722008-11-10010 November 2008 Relief Request Nos. 18 and 36, Proposed Alternatives for the Third Interval 10-Year Inservice Inspection Program Interval ML0825905562008-10-0202 October 2008 Relief Request No. 34 for Second 10-Year Inservice Inspection Interval ML0822506752008-07-11011 July 2008 American Society of Mechanical Engineers (ASME) Code, Section XI, Request for Approval of an Alternative Repair Method - Relief Request No. 39 ML0817002812008-06-0606 June 2008 Request for Extension to Complete the Final Confirmatory Analysis and Validation of Containment Sump Strainers Associated with NRC GL-04-002 ML0715600082007-06-21021 June 2007 Relief Request Nos. 36 and 37 Alternatives to Weld Overlay Requirements for Inservice Inspection ML0711400332007-05-16016 May 2007 Supplement to Relief Request No. 34 Request to Extend the Second 10-Year Inservice Inspection Program Interval for Reactor Vessel Weld Examinations ML0629302082006-11-0303 November 2006 Relief Request No. 32, Risk-Informed Inservice Inspection Program ML0626401992006-10-0404 October 2006 Relief Request No. 35 Request to Extend the Second 10-Year Inservice Inspection Program Interval for Reactor Vessel Visual Examinations ML0624905132006-09-20020 September 2006 Relief, Request to Extend the Second 10-Year Inservice Inspection Program Interval ML0623003332006-09-12012 September 2006 Relief Request No. 31, Revision 1, Proposed Alternative Repair for Reactor Coolant System Hot Leg Alloy 600 Small-Bore Nozzles ML0615702092006-05-26026 May 2006 Additional Information to Support the Request to Extend the Second 10-Year, American Society of Mechanical Engineers Section IX, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations - Relief Request No. 34 ML0610803942006-04-0707 April 2006 Relief, Request to Use a Later Edition and Addenda of the ASME Boiler and Pressure Vessel Code, Section XI for Repair/Replacement Activities at PVNGS in Accordance with 10 CFR 50.55a(g)(4)(iv) ML0518901452005-06-28028 June 2005 Revision 1 to 10 CFR 50.55a(a)(3)(i) Request for Alternatives to 10 CFR 50.55a(c) Requirement to Comply with ASME Section III, Subsection NB-1120, Temperature Limits, for a Portion of the Plant Pressurizer That Was Subjected To. ML0512901232005-05-0505 May 2005 Relief, Relief Request No. 31 Proposed Alternative Repair for Reactor Coolant System Hot Leg Alloy 600 Small-Bore Nozzles ML0434201672004-11-24024 November 2004 Relief Request No. 25 for Palo Verde, Unit 2 - Relaxation from First Revised NRC Order EA-03-009 - Additional Analysis Information for Control Element Drive Mechanism (CEDM) Nozzles ML0432706262004-11-19019 November 2004 Palo Verde - Correction to Approval Letter for Relief Request No. 30 ML0328705392003-10-0909 October 2003 Relief, Inservice Inspection Program Surface Examination Requirements, (TACs MC0830, MC0831, and MC0832) ML0321105422003-07-30030 July 2003 Relief Request 23 Alternative to Temper Bead Welding Requirements for Inservice Inspection Program (Tac. MB8973, MB8974 and MB8975) ML0314000512003-05-15015 May 2003 CFR 50.55a Alternative Repair Request for the Second 10-Year Interval of the Inservice Inspection Program: Relief Request 23, Pressurizer Heater Sleeves ML0302902122003-01-27027 January 2003 Relief, Requirements of American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code (Code) Concerning Use of Electrical Discharge Machining (Edm), MB6439, MB6440 & MB6441 ML0302801692003-01-24024 January 2003 Mechanical Nozzle Seal Assembly Type 2 Code Replacement Request for Relief from 10 CFR 50.55a ML0235200682002-12-11011 December 2002 Resubmittal of 10 CFR 50.55a Alternative Repair Request for the Second 10-Year Interval of the Inservice Inspection Program (Relief Request 18) ML0227404992002-09-25025 September 2002 CFR 50.55a Alternative Repair Request for the Second 10-Year Interval of the Inservice Inspection Program (Relief Request 22) 2022-05-12
[Table view] Category:Inservice/Preservice Inspection and Test Report
MONTHYEARML24193A3442024-07-11011 July 2024 Fourth 10-Year Interval, Second Period Owner’S Activity Report Number 3R24 ML23188A1872023-07-0707 July 2023 Fourth 10-Year Interval, Second Period Owners Activity Report Number 2R24 ML23013A0912023-01-13013 January 2023 Fourth 10-Year Interval, Second Period Owners Activity Report Number 3R23 ML22189A1432022-07-0808 July 2022 Fourth 10-Year Interval, First Period Owner'S Activity Report Number 1R23 ML22018A2022022-01-18018 January 2022 Fourth 10-Year Interval, First Period Owner'S Activity Report Number 2R23 ML21341A5522021-12-0303 December 2021 Snubber Program for the Fourth 10-Year Testing Interval ML21292A1842021-10-19019 October 2021 Nd Refueling Outage Steam Generator Tube Inspection Report ML21197A2052021-07-16016 July 2021 Fourth 10-Year Interval, First Period Owner'S Activity Report Number 3R22 ML21047A4972021-02-16016 February 2021 Twenty-Second Refueling Outage Owner'S Activity, Report 1R22 ML20288A2982020-10-13013 October 2020 Refueling Outage 22 Steam Generator Inspection Report ML20203M2042020-07-17017 July 2020 Fourth 10-Year Interval, First Period Owner'S Activity Report Number 2R22 ML20054A2692020-02-19019 February 2020 Response to Request for Additional Information - Relief Request 64 - Unit 1 Impractical Examinations for the Third 10-Year Inservice Inspection Interval ML20028D5402020-01-24024 January 2020 Owner'S Activity Report Number U3R21 ML19291F5762019-10-18018 October 2019 Generator (Pvggs), Unit 1 - Steam Generataor Tube Inspection Ruport ML19205A2642019-07-24024 July 2019 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U1R21 ML19198A3402019-07-17017 July 2019 Relief Request 64 - Unit 1 Impractical Examinations for the Third 10-Year Inservice Inspection Interval ML19179A3312019-06-28028 June 2019 Response to Request for Additional Information - Relief Request 63 - Unit 3 Impractical Examinations for the Third 10-Year Inservice Inspection Interval ML19017A3452019-01-17017 January 2019 Fourth 10-Year Interval Pump and Valve Inservice Testing Program ML19010A3072019-01-10010 January 2019 Relief Request 63 -Unit 3 Impractical Examinations for the Third 10-Year Inservice Inspection Interval ML18306A9992018-11-0202 November 2018 Steam Generator Tube Inspection Report ML18277A2972018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update, Part 1 of 7 ML18277A3122018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update, Part 4 of 7 ML18277A3032018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update, Part 2 of 7 ML18277A3102018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update. (Part 3 of 7) ML18277A3232018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update, Part 7 of 7 ML18277A3212018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update, Part 6 of 7 ML18277A3192018-09-21021 September 2018 Third 10-Year Interval Inservice Inspection Program Update, Part 5 of 7 ML18102B6842018-04-12012 April 2018 Third 10-Year Interval, Third Period: Owner'S Activity Report Number 1R20 Correction ML18093A3352018-03-30030 March 2018 Palo Verde, Units 1, 2 and 3 - Fourth 10-Year Interval Pump and Valve Inservice Testing Program ML18030B1382018-01-26026 January 2018 Third 10-Year Interval, Third Period: Owner'S Activity Report Number 1R20 ML17270A4242017-09-26026 September 2017 Transmittal of 20th Refueling Outage Steam Generator Tube Inspection Report ML17181A5202017-06-30030 June 2017 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U2R20 ML17248A5232017-05-30030 May 2017 3INT-ISI-1, Revision 5, Third Inspection Interval Inservice Inspection Program Summary Manual, Unit 1, Enclosure 1 to 102-07551-MDD/MSC ML17248A5252017-05-26026 May 2017 3INT-ISI-2, Revision 6, Third Inspection Interval Inservice Inspection Program Summary Manual, Unit 2, Enclosure 2 to 102-07551-MDD/MSC ML17054D6872017-02-23023 February 2017 Fourth 10-Year Interval Pump and Valve Inservice Testing Program Relief Requests GRR-01, GRR-02, PRR-01, PRR-02, PRR-03, PRR-04, PRR-05, and PRR-06 ML16308A2322016-10-26026 October 2016 Steam Generator Tube Inspection Report - Refueling Outage 19 ML16180A4462016-06-24024 June 2016 Third 10-Year Interval, Third Period: Owner'S Activity Report Number U1R19 ML16015A4542016-01-15015 January 2016 Submittal of Owner'S Activity Report (Form OAR-1) Refueling Outage 19 (U2R19) ML15267A2692015-09-23023 September 2015 Steam Generator Tube Inspection Report ML14035A4042013-08-29029 August 2013 3INT-ISI-3, Rev. 3, Palo Verde, Unit 3, 3rd Interval: Inservice Inspection Program Summary Manual ML14035A4022013-08-28028 August 2013 3INT-ISI-2, Rev. 4, Palo Verde, Unit 2, 3rd Interval: Inservice Inspection Program Summary Manual ML14035A4002013-08-22022 August 2013 INT-ISI-1, Rev. 3, 3rd Interval, Inservice Inspection Program Summary Manual ML13182A0152013-06-26026 June 2013 Third 10-Year Interval, Second Period: Owner'S Activity Report Number U1R17 ML14035A3992013-05-31031 May 2013 INT-ISI-1, Rev. 2, 3rd Inspection Interval, Inservice Inspection Program Summary Manual ML14035A4012013-05-31031 May 2013 INT-ISI-2, Rev. 3, 3rd Inspection Interval, Inservice Inspection Program Summary Manual ML14035A4032013-05-31031 May 2013 3INT-ISI-3, Rev. 2, Palo Verde, Unit 3, 3rd Interval: Inservice Inspection Program Summary Manual ML12242A2492012-08-0707 August 2012 Steam Generator Tube Inspection Report ML11287A0142011-10-0404 October 2011 Revised Third 10 Year Interval Inservice Inspection Plans ML11136A1062011-05-0505 May 2011 Submittal of Steam Generator Tube Inspection Report ML0916300622009-06-0404 June 2009 Submission of Relief Request 36 Revision 1 to the American Society of Mechanical Engineers Section Xl, Inservice Inspection Program Third Interval 2024-07-11
[Table view] Category:Letter
MONTHYEARML24262A0972024-09-23023 September 2024 Notification of Post-Approval Site Inspection for License Renewal and Request for Information Inspection (05000529/2024011) ML24241A2542024-08-28028 August 2024 Inservice Inspection Request for Information ML24241A2782024-08-28028 August 2024 License Amendment Request to Revise the Technical Specifications 3.5.1 and 3.5.2 Safety Injection Tank Pressure Bands, and to Use GOTHIC Code ML24240A2682024-08-27027 August 2024 Transmittal of Technical Specification Bases Revision 79 IR 05000528/20240052024-08-22022 August 2024 Updated Inspection Plan for Palo Verde Nuclear Generating Station - Units 1, 2, and 3 (Report 05000528/2024005, 05000529/2024005, 05000530/2024005) 05000530/LER-2024-001, Inoperable Boron Dilution Alarm System(Bdas) with Technical Specification Violation2024-08-21021 August 2024 Inoperable Boron Dilution Alarm System(Bdas) with Technical Specification Violation ML24208A0612024-08-20020 August 2024 Issuance of Amendment Nos. 224, 224, and 224 Regarding Revision to Technical Specifications 3.5.1, 3.5.2 and 3.6.5 IR 05000528/20244042024-08-0808 August 2024 Cybersecurity Inspection Report 05000528/2024404, 05000529/2024404 and 05000530/2024404 ML24213A3292024-07-31031 July 2024 Transmittal of Relief Request (RR) No. 72: Re-Submittal of RR-39 ML24213A3232024-07-31031 July 2024 Transmittal of Relief Request (RR) No. 71: Re-Submittal of RR-30 IR 05000528/20240022024-07-29029 July 2024 Integrated Inspection Report 05000528/2024002 and 05000529/2024002 and 05000530/2024002 ML24173A3302024-07-24024 July 2024 Pressurizer Surge Line Inspection Program ML24159A4702024-07-17017 July 2024 Issuance of Amendment Nos. 223, 223, and 223 Revision to Technical Specifications 3.5.1 and 3.5.2 Using Risk Informed Process for Evaluations ML24198A0662024-07-16016 July 2024 Program Review - Simulator Testing Methodology ML24193A3442024-07-11011 July 2024 Fourth 10-Year Interval, Second Period Owner’S Activity Report Number 3R24 ML24129A0522024-07-0303 July 2024 Review of the Spring 2023 Steam Generator Tube Inspection Report IR 05000528/20240042024-06-25025 June 2024 Notification of Inspection (NRC Inspection Report 05000528/2024004, 05000529/2024004, 05000530/2024004) ML24177A3222024-06-25025 June 2024 Invalid Specified System Actuation of Train B Emergency Diesel Generator ML24177A3212024-06-25025 June 2024 Independent Spent Fuel Storage Installation, Registration of Dry Spent Fuel Transportable Storage Canisters Identification Numbers AMZDFX180, AMZDFX181, AMZDFX182 Vertical Concrete Cask Identification Nu ML24170A9962024-06-18018 June 2024 Response to Second Request for Additional Information to Revise Technical Specifications (TS) 3.5.1, Safety Injection Tanks (Sits) – Operating, TS 3.5.2, Safety Injection Tanks (Sits) – Shutdown a IR 05000528/20243012024-06-17017 June 2024 NRC Examination Report 05000528/2024301; 05000529/2024301; 05000530/2024301 ML24129A2062024-06-14014 June 2024 Issuance of Amendment Nos. 222, 222, and 222 Revision to Technical Specifications to Adopt TSTF-266-A ML24157A3392024-06-0505 June 2024 Inoperable Boron Dilution Alarm System(Bdas) with Technical Specification Violation ML24159A0262024-06-0303 June 2024 Annual Report of Guarantee of Payment of Deferred Premium ML24144A2772024-05-23023 May 2024 Valid Specified System Actuations of Unit 2 Train B Emergency Diesel Generator and Train B Auxiliary Feedwater ML24135A2482024-05-14014 May 2024 Response to Second Request for Additional Information to Proposed Method to Manage Environmentally Assisted Fatigue for the Pressurizer Surge Line ML24164A2582024-05-0909 May 2024 10-PV-2024-04 Post-Exam Comments ML24129A1482024-05-0707 May 2024 and Independent Spent Fuel Storage Installation Registration of Dry Spent Fuel Storage Casks with Applied Changes ML24128A2702024-05-0707 May 2024 Docket Nos. Stn 50-528/529/530 - Response to NRC Regulatory Issue Summary (RIS) 2024-01, Preparation and Scheduling of Operator Licensing Examinations IR 05000528/20240012024-05-0202 May 2024 Independent Spent Fuel Storage Installation, Integrated Inspection Report 05000528/2024001, 05000529/2024001, 05000530/2024001, 07200044/2024001, and Exercise of Enforcement Discretion ML24119A0022024-04-26026 April 2024 2023 Annual Environmental Operating Report ML24116A2082024-04-24024 April 2024 Independent Spent Fuel Storage Installation, Annual Radioactive Effluent Release Report 2023 IR 05000528/20244012024-04-22022 April 2024 Security Baseline Inspection Report 05000528/2024401 and 05000529/2024401 and 05000530/2024401 (Cover Letter) ML24109A0712024-04-22022 April 2024 NRC Initial Operator Licensing Examination Approval 05000528/2024301, 05000529/2024301, and 05000530/2024301 ML24112A0012024-04-19019 April 2024 Core Operating Limits Report, Revision 32 ML24108A1982024-04-16016 April 2024 Independent Spent Fuel Storage Installation - Registration of Dry Spent Fuel Storage Casks with Applied Changes Authorized by an Amended Certificate of Compliance ML24103A2482024-04-12012 April 2024 Emergency Core Cooling System Performance Evaluation Models, 10 CFR 50.46(a)(3)(ii) Annual Report for 2023 ML24131A0972024-04-10010 April 2024 Annual Radiological Environmental Operating Report 2023 ML24096A2202024-04-0505 April 2024 Transmittal of Technical Specification Bases Revision 78 ML24032A1542024-04-0303 April 2024 Exemption from Select Requirements of 10 CFR Part 73 (Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting) ML24089A1892024-03-28028 March 2024 Present Levels of Financial Protection ML24068A2522024-03-0808 March 2024 License Amendment Request to Revise Technical Specifications 3.5.1 and 3.5.2 Using Risk-Informed Process for Evaluations ML24066A0472024-03-0606 March 2024 Response to Request for Additional Information to Revise Technical Specifications (TS) 3.5.1, Safety Injection Tanks (Sits) – Operating, TS 3.5.2, Safety Injection Tanks (Sits) – Shutdown and TS IR 05000528/20240122024-03-0404 March 2024 Age-Related Degradation Inspection Report 05000528/2024012, 05000529/2024012 and 05000530/2024012 IR 05000528/20230062024-02-28028 February 2024 Annual Assessment Letter for Palo Verde Nuclear Generating Station, Units 1, 2 and 3 Report 05000528/2023006 and 05000529/2023006 and 05000530/2023006 ML24053A3972024-02-22022 February 2024 License Renewal - Alloy 600 Management Program Plan ML24047A3632024-02-16016 February 2024 Independent Spent Fuel Storage Installation - Registration of Dry Spent Fuel Transportable Storage Canisters Identification Numbers AMZDFX177, AMZDFX178, AMZDFX179 Vertical Concrete Cask Identification No. AMZDNE179, AMZDNE178 IR 05000528/20230042024-02-12012 February 2024 Integrated Inspection Report Report 05000528/2023004 and 05000529/2023004 and 05000530/2023004 ML24019A2012024-01-19019 January 2024 Fourth 10-Year Interval, Second Period Owner’S Activity Report Number 1R24 2024-09-23
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10 CFR 50.55a(a)(3)(i)
David Mauldin Vice President Mail Station 7605 Palo Verde Nuclear Nuclear Engineering Tel: 623-393-5553 PO Box 52034 Generating Station and Support Fax: 623-393-6077 Phoenix, Arizona 85072-2034 102-05503-CDM/SAB/RJR May 26, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
References:
- 1. APS letter 102-05486-CDM/SAB/RJR, Request to Extend the Second 10-Year, American Society of Mechanical EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations - Relief Request No. 34, dated May 4, 2006.
- 2. Letter from Nuclear Regulatory Commission to Westinghouse Electric Company, "Summary of teleconference with the Westinghouse Owners Group regarding potential one cycle relief of reactor pressure vessel shell weld inspections at pressurized water reactors related to WCAP-1 6168-NP, 'Risk-Informed Extension of Reactor Vessel In-Service Inspection Intervals," dated January 27, 2005.
Dear Sirs:
SUBJECT:
Palo Verde Nuclear Generating Station (PVNGS)
Units 2 and 3 Docket Nos. STN 50-5291530 Additional Information to Support the Request to Extend the Second 10-Year, American Society of Mechanical EngineersSection XI, Inservice Inspection Program Interval for Reactor Vessel Weld Examinations -Relief Request No. 34 As a result of conversations with the NRC staff, Arizona Public Service Company (APS) is providing additional information in support of its May 4, 2006 request (Reference 1) to use an alternative to the requirements of the ASME Boiler and Pressure Vessel Code, Section X1, Paragraph IWB-2412, Inspection Program B, for PVNGS Units 2 and 3.
In Reference 2 that the staff agreed to licensees submitting relief requests for a one cycle extension of the reactor vessel weld examinations to provide additional time for completing evaluations and staff review associated with Reference 1, the Westinghouse Owners Group Topical Report, WCAP - 16168.
NRC Document Control Desk Page 2 Relief Request No. 34, Additional Information APS' technical justification to extend PVNGS's second inspection interval performance of Category B-A and B-D examinations by one fuel cycle provided in Reference 1 was not complete in that the additional information requested by the NRC from other licensees was not contained in APS' original request. This additional information is being provided in the enclosure to this letter and addresses the accident sequences identified by the NRC during its re-evaluation of the risk from pressurized thermal shock (PTS).
This letter contains no new commitments and no revisions to existing commitments.
APS is also changing the date by which approval of this request is needed to October 1, 2006, to support activities for the fall 2006 2R13 refueling outage.
If you have any questions about this change, please telephone Thomas N. Weber at (623) 393-5764.
Sincerely, CDM/SAB/RJR/gt
Enclosure:
Relief Request No. 34 - Additional Information Addressing the Accident Sequences Identified by the NRC during its Re-evaluation of the Risk from Pressurized Thermal Shock (PTS).
cc: B. S. Mallett NRC Region IV Regional Administrator M. B. Fields NRC NRR Project Manager G. G. Warnick NRC Senior Resident Inspector for PVNGS
ENCLOSURE Relief Request No. 34 - Additional Information Addressing the Accident Sequences Identified by the NRC during its Re-evaluation of the Risk from Pressurized Thermal Shock (PTS)
Enclosure - RR 34 - Additional Information NRC QUESTION:
You stated that the technical justification for your request was consistent with the guidance provided in a January 27, 2005, letter from the NRC to Westinghouse Electric Company (Summary of Teleconference with the Westinghouse Owners Group Regarding Potential One Cycle Relief of Reactor Pressure Vessel Shell Weld Inspections at Pressurized-Water Reactors Related to WCAP-16168-NP, "Risk Informed Extension of Reactor Vessel In-Service Inspection Intervals"). Item number six of this guidance is repeated below:
The licensee could then provide a discussion of how, based on its plant operational experience, fleet-wide operational experience, and plant characteristics, the likelihood of an event (in particular, a significant pressurized thermal shock event) over the next operating cycle which could challenge the integrity of the reactor vessel pressure vessel (RPV), if a flaw was present, is very low.
Section 5.5 of your submittal includes general statements indicating that the likelihood of pressurized thermal shock (PTS) events is small and briefly describes APS operating procedures that provide actions to avoid, or limit thermal shock to the reactor pressure vessel.
The NRC staff is re-evaluating the risk from PTS events in a study done to develop a technical basis for revising Title 10 of the Code of FederalRegulations, Part 50, Section 61 (10 CFR 50.61). Although the NRC staff has not yet completed its evaluation, the current results indicate that the following three types of accident sequences cause the more severe PTS events and thereby dominate the risk. Please describe the characteristics of your plant (design and operating procedures) that provide assurance that the likelihood of a severe PTS event over the next operating cycle which could challenge the integrity of the RPV, if a flaw was present, is very low.
Sequence 1:
Any transient with reactor trip followed by one stuck-open pressurizer safety relief valve that re-closes after about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Severe PTS events also require the failure to properly control high-head injection.
Sequence 2:
Large loss of secondary steam from steam line break or stuck-open atmospheric dump valves. Severe PTS events also require the failure to properly control auxiliary feedwater flow rate and destination (e.g., away from affected steam generators) and failure to properly control high pressure injection.
Page 1
Enclosure - RR 34 - Additional Information Sequence 3:
Four to nine-inch loss-of-coolant accidents. Severity of PTS event depends on break location (worst location appears to be in the pressurizer surge line) and primary injection systems flow rate and water temperature.
APS Response:
As discussed in Section 5.5 of APS' May 4, 2006 request, PVNGS Units 2 and 3 have implemented emergency operating procedures (EOP) and operator training to prevent the occurrence of PTS events. Consistent with the Combustion Engineering (CE)
Emergency Response Guidelines (ERGs), the PVNGS EOPs allow operators to identify the onset of PTS conditions and provide the steps required to mitigate any cold pressurization challenge to RV integrity. The basic PTS mitigation strategy of the PVNGS EOPs involves 1) termination of the primary system cool down, 2) termination of emergency core cooling system flow (if proper criteria are met), 3) depressurization of the primary system, 4) establishment of stable primary system conditions in the normal operating range, and 5) implementation of a thermal "soaking" period prior to any cool down outside of the normal operating region.
Sequence 1:
Any transient with reactor trip followed by one stuck-open pressurizer safety relief valve that re-closes after about I hour. Severe PTS events also require the failure to properly control high-head injection.
Initially, the control room personnel complete procedure 40EP-9EO01, "Standard Post Trip Actions," (SPTAs). This procedure is used for any event which actuates or requires a reactor trip. It is intended that the operator check each Safety Function and perform the Contingency Actions if necessary. The crew would then enter 40EP-9EO03, "Loss of Coolant Accident," (LOCA). The goals of this procedure are to mitigate the effects of a LOCA, to isolate the break (if possible), and to establish either long term cooling using the safety injection system or the shutdown cooling system. After some verification and notification steps, the crew reaches Step 23 within a few minutes. Steps 23 through 25 include the following guidance:
- 23. Perform the following:
- a. PERFORM Appendix 5, RCS and PZR Cooldown Log.
- b. Cooldown the Steam Generators using the SBCS.
b.1 Cooldown the Steam Generators using the ADVs by ONE of the following:
- Operation from the Control Room
- Appendix 18, Local ADV Operation Page 2
Enclosure - RR 34 - Additional Information
- 24. IF steaming to atmosphere, THEN inform Radiation Protection and the RMS Technician.
- 25. Depressurize the RCS to less than 385 psia [385 psia] by performing the following:
- a. Operate Main or Auxiliary Pressurizer spray and PERFORM Appendix 6, Spray Valve Actuation Data Sheet.
- b. IF Safety Injection throttle criteria are met, THEN control ANY of the following to lower RCS pressure.
- Charging and letdown flow
" HPSI flow Additionally, Step 27 includes the following guidance:
- 27. IF at least one HPSI Pump is operating, AND ALL of the following conditions exist:
- RCS is 24°F or more subcooled
- Pressurizer level is greater than 10% and NOT lowering
- At least one Steam Generator is available for RCS heat removal with level being maintained within or being restored to 45 - 60% NR
- RVLMS indicates RVUH level is 16% or more THEN throttle HPSI flow or stop the HPSI Pumps one pump at a time.
The steps listed above initiate an RCS cool down and depressurization and allows throttling or stopping high pressure safety injection (HPSI) flow as needed. Flow requirements to maintain the core covered and cooled will decrease as RCS pressure is lowered, or the pressurizer safety relief valve reseats during the cool down. Throttling and/or stopping HPSI prevents or minimizes the magnitude of re-pressurization of the RCS, thereby precluding PTS.
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Enclosure - RR 34 - Additional Information Sequence 2:
Large loss of secondary steam from steam line break or stuck-open atmospheric dump valves. Severe PTS events also require the failure to properly control auxiliary feedwater flow rate and destination (e.g., away from affected steam generators) and failure to properly control high pressure injection.
Initially, the control room personnel complete procedure 40EP-9EO01, "Standard Post Trip Actions." The Operating staff would then enter 40EP-9EO05, "Excess Steam Demand," (ESD). The goals of this procedure are to mitigate the effects of an ESD, maintain the plant in hot standby, or hot shutdown (if the break has been isolated), or to establish Shutdown Cooling System entry conditions while minimizing radiological releases to the environment and maintaining adequate core cooling. 40EP-9EO05 includes the following guidance in Step 14 which is performed after isolating the most affected Steam Generator, including stopping auxiliary feedwater to the faulted SG:
- 14. Stabilize RCS temperature using the lowest Tc by performing the following:
- a. Maintain Tc within the P/T limits.
REFER TO Appendix 2, Figures
- b. Steam the least affected Steam Generator using ANY of the following:
"SBCS
- ADVs from the Control Room
- Appendix 18, Local ADV Operation Stabilizing Tcold within the Pressure/Temperature (P/T) limits precludes PTS. EOP 40EP-9EO05 also contains the guidance for throttling HPSI when throttle criteria are met (same as Step 27 from LOCA). EOP 40EP-9EO1 0, "Standard Appendices,"
contains all of the figures, tables, charts, graphs and sub-procedures associated with performance of the EOPs.
Appendix 2 of 40EP-9EO1 0 shows the acceptable areas of operation and delineates where PTS becomes a concern (the 200 degree subcooled line).
Appendix 2 is provided below:
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Enclosure - RR 34 - Additional Information RCS Press Terrp Lirrdts Normal CTMT Conditions 100 lr MGo:*
- uF ii nubSd Yoo i I! I 2000 IVI
$S0 psla trarskion line QSPDS no longer
-- -r-- -
usd -----
0 I~ILt 50 L-.
100 I
150 200 250 300 350 400 450 500 850 600 RCS Temperature (Th F)
Forced Circulation - Th indicaton used Naluru Circulaon - REP CET used Sequence 3:
Four to nine-inch loss-of-coolant accidents. Severity of PTS event depends on break location (worst location appears to be in the pressurizer surge line) and primary injection systems flow rate and water temperature.
The response to this sequence would be the same as Sequence 1.
Additionally, as part of a recent power uprate (PUR) amendment request, APS performed fluence calculations using the existing analysis of record (AOR) at the 4200 MWt power level. An out-in type of fuel loading was assumed, however, the proposed PUR was for 3990 MWt and the loading pattern has been low leakage for a number of cycles. Both conditions are conservative and the AOR bounds the values calculated for the PUR. With the issuance of Amendment No. 157 to Facility Operating License Nos.
NPF-51, and NPF-74, dated November 16, 2005, the NRC acknowledged that the P-T curves currently approved for Units 2 and 3 were valid for 32 effective full power years (EFPY). Since Units 2 and 3 are estimated to be at only 17.3 and 17.6 EFPY respectively at the end of operating cycle 14, sufficient margin exists until the examinations are performed during 14 refueling outages for each unit.
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