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Category:Code Relief or Alternative
MONTHYEARML23256A0882023-09-25025 September 2023 Issuance of Alternative to Steam Generator Welds ML23151A3482023-05-30030 May 2023 Duke Fleet - Request for Additional Information Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML23118A0762023-05-0101 May 2023 Approval for Use of Specific Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RA-22-0257, Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)2023-02-17017 February 2023 Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1) ML22096A0032022-11-18018 November 2022 McGuire Nuclear Station and Shearon Harris Nuclear Power Plant Authorization of RA-19-0352 Regarding Use of Alternative for RPV Head Closure Stud Examinations ML22213A2532022-08-31031 August 2022 Authorization and Safety Evaluation for Alternative Request RA-21-0144 to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method ML22007A3452022-01-11011 January 2022 Proposed Alternative to Use the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-885 ML21029A3352021-02-16016 February 2021 Request to Use a Provision of a Later Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI RA-19-0026, Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 and 10 (18-GO-001)2019-02-11011 February 2019 Response to Request for Additional Information (RAI) Regarding Proposed Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 and 10 (18-GO-001) RA-18-0180, Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 22018-11-12012 November 2018 Supplement to Relief Request for an Alternative to the Depth Sizing Qualification Requirement of Appendix Viii, Supplements 2 ML18016A1782018-01-24024 January 2018 Request for Relief No. 17-CN-001, Regarding Category B-J Pressure Retaining Welds for the Third 10-Year Inservice Testing Interval (CAC Nos. MF9807 and MF9808; EPID L-2017-LLA-0032) ML17331A0862017-12-26026 December 2017 Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads ML17104A1102017-04-26026 April 2017 Relief Request CN Grr 01, Alternative to the Testing Frequencies in the American Society of Mechanical Engineers Operation and Maintenance Code, by Adoption of Code Case OMN-20, Inservice Test Frequencies (CAC Nos. MF9595-6) CNS-17-020, Application to Adopt ASME Code Case OMN-20 (CN-GRR-01) for Pump and Valve Inservice Testing2017-04-11011 April 2017 Application to Adopt ASME Code Case OMN-20 (CN-GRR-01) for Pump and Valve Inservice Testing ML15301A5442015-11-10010 November 2015 Proposed Relief Request 15-CN-001, Alternative Repair of a Main Steam System Braided Flex-Hose CNS-15-062, Relief Request Serial Number 15-CN-001, Proposed Alternative Repair for Main Steam System Braided Flex-Hose - Submitted Pursuant to 10 CFR 50.55a(z)(2)2015-07-16016 July 2015 Relief Request Serial Number 15-CN-001, Proposed Alternative Repair for Main Steam System Braided Flex-Hose - Submitted Pursuant to 10 CFR 50.55a(z)(2) ML15161A3762015-06-19019 June 2015 Proposed Relief Request 14-CN-002, Alternative to American Society of Mechanical Engineers (ASME) Section XI Requirements for Class 3 Buried Piping ML15124A6992015-05-0808 May 2015 (Catawba 1 and 2) - Request to Use a Later Edition of the ASME Boiler and Pressure Vessel Code, Section X1, for Inservice Inspection Activities ML14295A5322014-10-30030 October 2014 Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME) Section IX Volumetric Examination Requirements ML14079A5462014-03-26026 March 2014 Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds ML11187A0012011-07-14014 July 2011 Relief 10-CN-001, Alternative Requirements for Temporary Acceptance of a Through-Wall Flaw in Boric Acid Tank Nozzle Weld ML1012501702010-05-0303 May 2010 Withdrawal of Request for Relief Number 09-CN-001, Limited Weld Examinations During the End-of-Cycle 17 Refueling Outage ML0912401562009-05-27027 May 2009 Relief 06-CN-003 for Use of Polyethylene Material in Buried Service Water Piping ML0902301722009-01-16016 January 2009 Response to NRC Requests for Additional Information and Request for Relief (Alternative) Number 08-CN-002 - Reactor Vessel Hot Leg Nozzle-to-Safe End Weld Full Structural Weld Overlays ML0831506302008-11-0606 November 2008 Transmittal of Request for Relief Number 08-CN-002, Reactor Vessel Hot Leg Nozzle-to-Safe End Weld Full Structural Weld Overlays (Fswols) ML0822403862008-09-12012 September 2008 Relief 06-CN-003, Use of Polyethylene Material in Nuclear Safety-Related Piping Applications ML0815607342008-07-0707 July 2008 Request for Relief 07-CN-001, Limited Weld Examinations During End-Of -Cycle 14 Refueling Outage ML0816000032008-07-0101 July 2008 Request for Relief 07-CN-002, Limited Weld Examinations During End-of-Cycle 14 Refueling Outage ML0731804472007-08-27027 August 2007 Request for Relief Number 07-CN-004 Limited Weld Examinations During the End-of-Cycle 16 Refueling Outage ML0706504172007-02-26026 February 2007 Submittal of Request for Relief Number 07-CN-003, Request for Relief to Allow Use of Alternate Requirements for Snubber Inspection & Testing ML0636000142007-01-0909 January 2007 Relief, Request to Use a Portion of a Letter Edition of ASME Code ML0632801882006-11-16016 November 2006 Revision to Relief Request 06-GO-001 ML0623900202006-09-25025 September 2006 Request for Relief 05-CN-004, Limited Weld Examinations During End-of-Cycle 15 Refueling Outage, MC8337, MC9171, MC9172, MC9173, MC9174, MC9175, MC9176, MC9177, MC9178, and MC9179) ML0626202612006-09-11011 September 2006 Relief Request 06-GO-001 Request for Additional Information ML0622304212006-09-0707 September 2006 Relief, Request for Relief for Snubber Visual Examination and Functional Testing Related to the Third 10-year Interval Inservice Inspection Program ML0634903992006-07-27027 July 2006 Relief Request 06-GO-001 ML0616603262006-07-20020 July 2006 Request for Use of Later Edition and Addenda of the ASME Code ML0616800102006-07-19019 July 2006 Regarding Request for Relief 05-CN-001, Limited Weld Examinations During End-of-Cycle 14 Refueling Outage ML0535503702006-04-0303 April 2006 Relief, O5-CN-003 Limited Weld Examination Coverage ML0517801882005-07-25025 July 2005 RR 04-CN-001, Volumetric Weld Inspection Requirements ML0519403672005-07-13013 July 2005 Relief Request MC3057, 04-CN-002, Limited Weld Examinations During End-of-Cycle 12 Refueling Outage ML0518901432005-06-28028 June 2005 Submittal of Third Ten-Year Interval Inservice Inspection Plan for Steam Generator Tubing ML0517801642005-06-23023 June 2005 RR 04-CN-004, Class 1 ISI Pressure Test Boundaries Requirement (Tac No. MC3276, MC3277) ML0510203642005-03-31031 March 2005 Relief, Request to Use a Later Edition and Addenda of the ASME Boiler and Pressure Vessel Code, Section XI for Repair/Replacement Activities at Catawba Nuclear Station, Unit 2 in Accordance with 10 CFR 50.55a(g) (4) (IV) ML0504105322005-03-17017 March 2005 RR, Alternative Code of Prv Nozzle Welds for Volumetic Examination RR-04-GO-001 ML0504904622005-02-17017 February 2005 Relief, RR No. 03-001, 10 CFR Section 50.55a (a)(3)(ii) for Second ISI Intervals ML0428106012004-10-20020 October 2004 Relief Request, Alternate to the American Society of Mechanical Engineers Section XI for Reactor Vessel Examinations RR-04-GO-001 ML0427400272004-09-30030 September 2004 Relief, Amended Relief Request 03-GO-010 Requesting Use of Subsequent Edition and Addenda of the ASME Code for Class Mc and CC Components (MC9674-MC9680) ML0427302912004-09-29029 September 2004 Request for Relief, 04-GO-006, for Use of an Alternate Code Case to Use Ultrasonic Techniques Equipment ML0420403942004-07-20020 July 2004 Request for Relief 03-004, Minimum Percentage Requirements for Second Period of the Second Inservice Inspection Interval 2023-09-25
[Table view] Category:Letter type:CNS
MONTHYEARCNS-18-008, License Amendment Request to Remove the Containment Valve Injection Water System Supply from Specified Containment Isolation Valves and to Exempt These CIVs from Type-C Local Leak Rate Testing2018-07-19019 July 2018 License Amendment Request to Remove the Containment Valve Injection Water System Supply from Specified Containment Isolation Valves and to Exempt These CIVs from Type-C Local Leak Rate Testing CNS-18-032, Refueling Outage 22 (C2R22) Inservice Inspection Report Summary Report2018-07-12012 July 2018 Refueling Outage 22 (C2R22) Inservice Inspection Report Summary Report CNS-18-031, Notice of Enforcement Discretion (NOED) Request2018-06-18018 June 2018 Notice of Enforcement Discretion (NOED) Request CNS-18-028, 2017 Annual Radiological Environmental Operating Report2018-05-14014 May 2018 2017 Annual Radiological Environmental Operating Report CNS-18-026, Transmittal of 2017/2018 10 CFR 50.59 Evaluation Summary Report2018-05-10010 May 2018 Transmittal of 2017/2018 10 CFR 50.59 Evaluation Summary Report CNS-18-023, (Cns), Units 1 and 2 - Response to NRC Requests for Additional Information (Rais) License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System2018-05-0808 May 2018 (Cns), Units 1 and 2 - Response to NRC Requests for Additional Information (Rais) License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System CNS-18-025, Submittal of 2017 Annual Radioactive Effluent Release Report2018-04-30030 April 2018 Submittal of 2017 Annual Radioactive Effluent Release Report CNS-18-022, Submittal of Emergency Plan Implementing Procedure2018-04-12012 April 2018 Submittal of Emergency Plan Implementing Procedure CNS-18-017, Catawba, Units 1 and 2, Submittal of Emergency Plan, Revision 18-1 and 18-22018-03-24024 March 2018 Catawba, Units 1 and 2, Submittal of Emergency Plan, Revision 18-1 and 18-2 CNS-18-015, Commitment Change Evaluation Report for 20172018-03-22022 March 2018 Commitment Change Evaluation Report for 2017 CNS-18-005, Technical Specification Bases Changes and Ufsar/Selected Licensee Commitment Changes2018-02-0808 February 2018 Technical Specification Bases Changes and Ufsar/Selected Licensee Commitment Changes CNS-18-006, Submittal of Special Report 2018-01 for the Loose Parts Monitoring System Non-Functional for Greater than 30 Days2018-01-23023 January 2018 Submittal of Special Report 2018-01 for the Loose Parts Monitoring System Non-Functional for Greater than 30 Days CNS-17-058, Supplement to the High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukus2017-12-28028 December 2017 Supplement to the High Frequency Supplement to Seismic Hazard Screening Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushi CNS-17-061, Third Ten-Year Inservice Inspection Plan, Limited Volumetric Examinations Response to NRC Requests for Additional Information2017-12-21021 December 2017 Third Ten-Year Inservice Inspection Plan, Limited Volumetric Examinations Response to NRC Requests for Additional Information CNS-17-056, Catawba Nuclear Station Units 1 and 2, Resubmittal of Emergency Plan, Revision 17-22017-11-0101 November 2017 Catawba Nuclear Station Units 1 and 2, Resubmittal of Emergency Plan, Revision 17-2 CNS-17-056, Catawba Nuclear Station Units 1 and 2, Submittal of Emergency Plan, Revision 17-22017-10-23023 October 2017 Catawba Nuclear Station Units 1 and 2, Submittal of Emergency Plan, Revision 17-2 CNS-17-053, Transmittal of 2016 Annual Radioactive Effluent Release Report - Supplement2017-10-12012 October 2017 Transmittal of 2016 Annual Radioactive Effluent Release Report - Supplement CNS-17-048, Supplemental Response to Preliminary White Finding, Apparent Violations, and NRC Inspection Report 05000414/2017012; Enforcement Action 17-1222017-09-29029 September 2017 Supplemental Response to Preliminary White Finding, Apparent Violations, and NRC Inspection Report 05000414/2017012; Enforcement Action 17-122 CNS-17-049, Submittal of Ufsar/Selected Licensee Commitment Changes Manual2017-09-28028 September 2017 Submittal of Ufsar/Selected Licensee Commitment Changes Manual CNS-17-051, Independent Spent Fuel Storage Installation (ISFSI) Registration of Spent Fuel Storage Cask2017-09-28028 September 2017 Independent Spent Fuel Storage Installation (ISFSI) Registration of Spent Fuel Storage Cask CNS-17-014, License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System, to Add a New Condition to Allow Single Pond Return Header Operation of the Nsws with a 30-Day Completion Time2017-09-14014 September 2017 License Amendment Request to Revise Technical Specification Section 3.7.8, Nuclear Service Water System, to Add a New Condition to Allow Single Pond Return Header Operation of the Nsws with a 30-Day Completion Time CNS-17-050, ISFSI - Registration of Spent Fuel Storage Cask2017-09-12012 September 2017 ISFSI - Registration of Spent Fuel Storage Cask CNS-17-047, Response to NRC Inspection Report 05000414/2017012 and Preliminary White Finding2017-08-29029 August 2017 Response to NRC Inspection Report 05000414/2017012 and Preliminary White Finding CNS-17-041, ISFSI - Registration of Spent Fuel Storage Cask2017-08-15015 August 2017 ISFSI - Registration of Spent Fuel Storage Cask CNS-17-040, Withdrawal of Inservice Testing Relief Request 17-CN-002 Regarding an Alternative Method to Establish New Reference Values for Testing Chemical and Volume Control Pump2017-08-14014 August 2017 Withdrawal of Inservice Testing Relief Request 17-CN-002 Regarding an Alternative Method to Establish New Reference Values for Testing Chemical and Volume Control Pump CNS-17-039, Relief Request 17-CN-002, Request for Alternative Method to Establish New Inservice Testing Values for a Chemical and Volume Control Pump2017-07-27027 July 2017 Relief Request 17-CN-002, Request for Alternative Method to Establish New Inservice Testing Values for a Chemical and Volume Control Pump CNS-17-036, Mid-Cycle Power Uprate Ascension Startup Report2017-06-29029 June 2017 Mid-Cycle Power Uprate Ascension Startup Report CNS-17-033, Technical Specification Bases Changes2017-06-27027 June 2017 Technical Specification Bases Changes CNS-17-029, Flood Hazard Mitigating Strategies Assessment Report Submittal2017-06-20020 June 2017 Flood Hazard Mitigating Strategies Assessment Report Submittal CNS-17-030, Catawba, Unit 1, Submittal of Core Operating Limits Report (Colr) for Cycle 24 Reload Core2017-06-20020 June 2017 Catawba, Unit 1, Submittal of Core Operating Limits Report (Colr) for Cycle 24 Reload Core CNS-17-024, Third Ten-Year Inservice Inspection Plan, Limited Volumetric Examinations2017-05-25025 May 2017 Third Ten-Year Inservice Inspection Plan, Limited Volumetric Examinations CNS-17-031, Response to NRC Requests for Additional Information End of Cycle 21 Refueling Outage Inservice Inspection Report and Steam Generator Inservice Inspection Summary Report2017-05-24024 May 2017 Response to NRC Requests for Additional Information End of Cycle 21 Refueling Outage Inservice Inspection Report and Steam Generator Inservice Inspection Summary Report CNS-17-027, Submittal of 2015/2016 10 CFR 50.59 Evaluation Summary Report2017-05-10010 May 2017 Submittal of 2015/2016 10 CFR 50.59 Evaluation Summary Report CNS-17-015, Catawba Nuclear Station, Units 1 and 2 - Submittal of 2016 Annual Radioactive Effluent Release Report2017-04-27027 April 2017 Catawba Nuclear Station, Units 1 and 2 - Submittal of 2016 Annual Radioactive Effluent Release Report CNS-17-021, Special Report 2017-01 for the Units 1 and 2 Loose Parts Monitoring System Non-Functional for Greater than 30 Days2017-04-20020 April 2017 Special Report 2017-01 for the Units 1 and 2 Loose Parts Monitoring System Non-Functional for Greater than 30 Days CNS-17-020, Application to Adopt ASME Code Case OMN-20 (CN-GRR-01) for Pump and Valve Inservice Testing2017-04-11011 April 2017 Application to Adopt ASME Code Case OMN-20 (CN-GRR-01) for Pump and Valve Inservice Testing CNS-17-008, Commitment Change Evaluation Report for 20162017-03-16016 March 2017 Commitment Change Evaluation Report for 2016 CNS-17-004, Response to NRC Request for Additional Information on License Amendment Request to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants.2017-01-26026 January 2017 Response to NRC Request for Additional Information on License Amendment Request to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants. CNS-17-002, End of Cycle 21 Refueling Outage, Lnservice Inspection Report and Steam Generator Lnservice Inspection Summary Report2017-01-0606 January 2017 End of Cycle 21 Refueling Outage, Lnservice Inspection Report and Steam Generator Lnservice Inspection Summary Report CNS-16-080, Submittal of Emergency Plan Implementing Procedure2016-12-20020 December 2016 Submittal of Emergency Plan Implementing Procedure CNS-16-077, Expedited Seismic Evaluation Process Closeout, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2016-12-0808 December 2016 Expedited Seismic Evaluation Process Closeout, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident CNS-16-066, Stations - Supplemental Information Regarding Reevaluated Seismic Hazard Screening and Prioritization Results - Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task..2016-10-20020 October 2016 Stations - Supplemental Information Regarding Reevaluated Seismic Hazard Screening and Prioritization Results - Response to NRC Request for Information, Per 10CFR50.54(f) Regarding Recommendation 2.1 of the Near-Term Task.. CNS-16-069, Submittal of Emergency Plan, Revision 16-22016-10-10010 October 2016 Submittal of Emergency Plan, Revision 16-2 CNS-16-068, Submittal of Emergency Plan Implementing Procedure Revisions2016-10-10010 October 2016 Submittal of Emergency Plan Implementing Procedure Revisions CNS-16-064, License Amendment Request to Adopt National Fire Protection and Revised Response to Probabilistic Risk Assessment Request for Additional Information 242016-09-0707 September 2016 License Amendment Request to Adopt National Fire Protection and Revised Response to Probabilistic Risk Assessment Request for Additional Information 24 CNS-16-061, Submittal of Design Basis Specification for the Plant Fire Protection Revision 252016-08-12012 August 2016 Submittal of Design Basis Specification for the Plant Fire Protection Revision 25 CNS-16-051, License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants, Response to NRC Request for Additional Information (RAI) an2016-08-0202 August 2016 License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants, Response to NRC Request for Additional Information (RAI) and CNS-16-049, ISFSI Registration of Spent Fuel Storage Cask2016-07-14014 July 2016 ISFSI Registration of Spent Fuel Storage Cask CNS-16-056, Transmittal of Emergency Plan Implementing Procedures, Including Revision 034 to RP/0/A/5000/006A and Revision 038 to RP/0/A/5000/00682016-07-13013 July 2016 Transmittal of Emergency Plan Implementing Procedures, Including Revision 034 to RP/0/A/5000/006A and Revision 038 to RP/0/A/5000/0068 CNS-16-050, Unit 2 - Ufsar/Selected Licensee Commitment Changes2016-07-12012 July 2016 Unit 2 - Ufsar/Selected Licensee Commitment Changes 2018-07-19
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Kelvin Henderson Vice President wo@ENERGY, Catawba Nuclear Station Duke Energy CNO1VP I 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 CNS-1 5-062 July 16, 2015 10 CFR 50.55a U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
Subject:
Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Unit 1 Docket Number 50-413 Relief Request Serial Number 15-CN-001, Proposed Alternative Repair for Main Steam System Braided Flex-Hose - Submitted Pursuant to 10 CFR 50.55a(z)(2)
References:
- 1. Letters from Duke Energy to NRC, same subject, dated March 19, 2015 (ADAMS Accession Number ML15082A074) and May 4, 2015 (ADAMS Accession Number ML15127A170).
- 2. Electronic mail from NRC to Duke Energy, Acceptance Review for 15-CN-001, dated April 20, 2015.
- 3. Conference call between Duke Energy and NRC, dated June 8, 2015.
The March 19, 2015 Reference 1 letter submitted Relief Request 15-CN-001 requesting approval to use an alternative repair for a Main Steam System leaking braided flex hose. The Reference 2 electronic mail requested additional information to enable the NRC to complete its acceptance review of the March 19, 2015 Reference 1 letter. The May 4, 2015 Reference 1 letter responded to the Reference 2 request. During the Reference 3 conference call, the NRC requested Duke Energy to clarify its plans for performing the subject alternative repair. The purpose of this letter is to provide the requested clarification.
This clarification is provided as enclosed revised Relief Request 15-CN-001. The body of the relief request is being resubmitted in its entirety. The relief request attachments have not changed from the original March 19, 2015 Reference 1 submittal; therefore, they are not being resubmitted with this relief request revision.
There are no regulatory commitments contained in this letter or its enclosure.
If you have any questions or require additional information, please contact L.J. Rudy at (803) 701-3084.
,uCR 1 www.duke-energy.com
U.S. Nuclear Regulatory Commission July 16, 2015 Page 2 Very truly yours, Kelvin Henderson Vice President, Catawba Nuclear Station LJR/s Enclosure
U.S. Nuclear Regulatory Commission July 16, 2015 Page 3 xc (with enclosure):
V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto III NRC Senior Resident Inspector Catawba Nuclear Station G.E. Miller (addressee only)
NRC Project Manager (Catawba)
U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738
Enclosure Duke Energy Carolinas, LLC Catawba Nuclear Station, Unit 1 Relief Request Serial No. 15-CN-001 Proposed Alternative Repair for Main Steam System Braided Flex-Hose, Submitted Pursuant to 10 CFR 50.55a(z)(2)
Relief Request 15-CN-001 Page 2 of 5 1.0 ASME Code Component Affected Catawba Nuclear Station, Unit 1 ASME Class 2 Level Instrument Flex-Hose.
The following information is applicable to this component:
System: Main Steam (SM)
Design Pressure: 1200 psia Design Temperature: 600°F Pipe Size and Material: 1/2"-Sch. 80 / SA376 TP304 Flex-Hose Size and Material: 1/"-2500# /ISA213TP304 Root Valves: 1/2"-1500# / SA1 82, Gr. F316 / globe valve (Model 09J-574)
The flex-hose is part of level instrument 1SMLS-5710. This degraded flex-hose is located downstream of the two stainless steel root valves on each side of the flex-hose as detailed in Catawba Unit 1 Instrument Weld Isometric drawing CNI-SM-1 571 (Attachment 2) and Weld Isometric drawing CN-1 SM-0082 (Attachment 3).
Instrument 1SMLS-5710 is located within the containment isolation boundary of the Main Steam 1A Containment Penetration (M 113) identified as Item No. 91 within UFSAR Table 6-77, Unit 1 Containment Isolation Valves Data.
2.0 Applicable Code Edition and Addenda
ASME Boiler and Pressure Vessel Code,Section XI, 1998 Edition with the 2000 Addenda 3.0 Applicable Code/Regqulatory Requirements 3.1 IWC-3516 specifies that acceptance standards for Examination Category C-H, All Pressure Retaining Components are in the course of preparation, and that the standards of IWB-3522 may be applied. Duke Energy has chosen to apply the acceptance standards of IWB-3522 to address leakage from this Class 2 component.
3.2 IWA-4400 specifies requirements for welding, brazing, defect removal, and installation of pressure retaining items.
3.3 IWA-4133 provides alternative requirements for repairs using mechanical clamping devices using Mandatory Appendix IX.
3.4 The ASME Boiler and Pressure Vessel Code, Section Xl, Appendix IX, Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary provides the requirements for using and designing mechanical clamping devices for piping pressure boundary.
4.0 Reason for Request
4.1 On August 12, 2014, a steam leak was discovered from a braided flex-hose part of level instrument 1SMLS-5710. This level switch is located in the Unit 1 Exterior Doghouse and controls valve 1SM-89 which dumps accumulated condensate to the Unit 1 main condenser as required. The flow diagram and weld isometric drawings showing this configuration are shown in Attachments 1, 2, and 3. The leaking ASME Class 2 flex-hose was isolated to comply with Catawba Selected Licensee Commitment (SLC) 16.5-5, "Structural Integrity-The structural integrity of the ASME Code Class 1, 2, and 3 components shall be maintained." Once isolated it was discovered that minor leakage still
Relief Request 15-CN-001 Page 3 of 5 occurred from the flex-hose due to valve seat leakage past the upstream root valves. To enable isolation of the leak to facilitate a code repair/replacement of the leaking flex-hose in accordance with IWA-4000, the alternative documented in this request is proposed. The proposed alternative consists of installing a mechanical clamping device using ASME Section Xl, Appendix IX and injecting sealant into the pipe between the two root valves to fully isolate the leaking component. Following sealant injection, the leaking flex-hose shall be removed and replacement pressure boundary material (pipe caps or plugs) shall be attached to the downstream side of the root valves to which the flex-hose had been connected. Because Appendix IX prohibits the use of mechanical clamping devices on portions of piping systems that form the containment boundary, relief is required to permit use of this mechanical clamping device to facilitate the isolation of these valves to replace the leaking flex-hose.
4.2 As defined in UFSAR Table 6-77, "Unit 1 Containment Isolation Valve Data",
the Main Steam containment isolation valves are not required to be leak rate tested. This is due to the Main Steam line being connected to the secondary side of the steam generator which is kept at a higher pressure than the primary side immediately after a LOCA occurs. Any leakage between the primary and secondary sides of the steam generator is directed inward to containment (i.e.,
main steam header pressure is maintained higher than peak containment pressure). This penetration is effectively sealed by steam header pressure against leakage from containment after a LOCA.
This Main Steam 1A Header leak has been evaluated against consequence during the postulated Steam Generator Tube Rupture (SGTR) Design Basis Accident. It was concluded that post SGTR radiation doses for this scenario would be bounded by those for the SGTR with a failed open Power Operated Relief Valve (PORV) on the ruptured Steam Generator. The presence of this steam leak does not invalidate any Safety Analysis calculations with regards to SGTR (i.e., an unanalyzed condition does not exist).
4.3 NRC Inspection Manual, Part 9900 Technical Guidance, Appendix C.12 Operational Leakage From ASME Code Class 1, 2, and 3 Components, states "The NRC staff does not consider through-wall conditions in components, unless intentionally designed to be there such as sparger flow holes, to be in accordance with the intent of the ASME Code or construction code and, therefore, would not meet code requirements, even though the system or component may demonstrate adequate structural integrity." The guidance provided in Part 9900 implies that the NRC does not accept that IWC-3000 of the ASME Code,Section XI allows through wall leakage in Class 2 components. Since a through-wall flaw in a flex-hose cannot be evaluated using an applicable and NRC endorsed code case, relief is required to comply with this guidance.
5.0 Proposed Alternative and Basis for Use 5.1 Proposed Alternative In lieu of the requirement of Appendix IX, Article IX-1000, General, (c) Clamping devices shall not be used on the following:
(2) portions of a piping system that forms the containment boundary
Relief Request 15-CN-001 Page 4 of 5 The following alternative requirements are proposed:
- 1. A non-code repair shall be performed to stop the leakage using a mechanical clamp, 1/8" NPT injection valve, and injection sealant. A mechanical clamp designed to the requirements of Appendix IX, Article Xl-3000 shall be installed followed by installation of an ASME Class 1 injection valve. After installing the injection valve, a 3/16" diameter hole shall be drilled in the pipe to facilitate sealant injection. After sealant injection is completed, the mechanical clamp and closed injection valve shall serve as part of the Class 2 pressure boundary until a code repair/replacement activity complying with IWA-4000 can be performed. A drawing of the mechanical clamp is provided in Attachment 4. Additionally, after verification that the leak has been fully isolated, the damaged braided flex-hose shall be removed and code compliant caps (or plugs) shall be installed on the end of the outboard root valves. These caps (or plugs) will also serve as part of the Class 2 pressure boundary until the flex-hose, affected root valves, and piping can be replaced.
- 2. The mechanical clamp, injection valve, and sealant injection may be used between one or both sets of root valves located upstream and downstream of the degraded flex-hose to fully isolate the leakage.
5.2 Basis for Proposed Alternative Hardship: A code-compliant repair cannot be performed without fully isolating and depressurizing the affected component. The root valves that are available to isolate this component are not leak-tight and are located upstream of the main steam isolation valves (MSIVs). Therefore, in order to isolate the affected component, a unit shutdown would be required to facilitate the repair. Therefore, the only way to perform a code-compliant repair in accordance with IWA-4000 would be to shut down the unit in order to depressurize the line and replace the affected components. Compliance with the specified requirement would result in hardship without a compensating increase in the level of quality and safety.
The proposed alternative to install an engineered clamp with an injection valve and inject sealant between the leaking root valves will enable the leaking component to be fully isolated and depressurized to permit removal of the leaking flex-hose and installation of pipe caps (or plugs) to restore the leak-tight integrity of the system. The piping between the leaking root valves has been evaluated and the structural and leak-tight integrity of this piping shall be maintained during the installation of the clamp, during leak injection, and during subsequent operation until a permanent repair/replacement activity can be performed.
6.0 Duration of Proposed Alternative The proposed alternatives to the ASME Code are applicable for the third 10-year Inservice Inspection (ISl) Interval at Catawba Nuclear Station, Unit 1.
The Catawba Unit 1 third Inservice Inspection Interval began on June 29, 2005 and is currently scheduled to end on June 29, 2016.
Use of the proposed alternative is requested until Code repair/replacement activities can be performed on the level instrument piping and flex-hose during refueling
Relief Request 15-CN-001 Page 5 of 5 outage 1EOC22 (Fall 2015) or during a forced outage of sufficient duration before refueling outage 1 EOC22.
7.0 References 7.1 1998 Edition through 2000 Addenda, ASME Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components."
7.2 US NRC Regulatory Issue Summary 2005-20, Rev. 1, Revision to NRC Inspection Manual Part 9900 Technical Guidance, "Operability Determinations
& Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety".