RS-23-108, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles: Difference between revisions

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{{#Wiki_filter:200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.55a RS-23-108 October 11, 2023 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50 456 and STN 50 457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50 454 and STN 50 455 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License Nos. DPR-18 NRC Docket Nos. 50-244
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==Subject:==
Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles
 
==References:==
: 1) Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Proposed Alternative for Examinations of Examination Category C-B Steam Generator Nozzle-to-Shell Welds and Nozzle Inside Radius Sections, dated September 1, 2021, ADAMS Accession No, ML21244A328.
: 2) Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles, dated December 14, 2021, ADAMS Accession No. ML21348A078.
: 3) Letter from D. Reddick (Constellation Energy Generation, LLC) to U.S.
Nuclear Regulatory Commission, Withdrawal - Proposed Alternatives Related to the Steam Generators, dated June 27, 2023, ADAMS Accession No. ML23178A169.
 
Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles October 11, 2023 Page 2 In the Reference 1 and 2 letters, Constellation Energy Generation, LLC (Constellation) requested Nuclear Regulatory Commission (NRC) approval of proposed alternatives for Braidwood Generating Station (Braidwood), Units 1 and 2, Byron Generating Station (Byron),
Units 1 and 2; Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2; and R. E.
Ginna Nuclear Power Plant (Ginna). Specifically, the proposed alternatives were related to extending the frequency of certain ASME Section XI, Class 1 and 2 Steam Generator examinations. In the Reference 3 letter, these proposed alternatives were withdrawn due to the continued discussions regarding performance monitoring requirements.
In accordance with 10 CFR 50.55a(z)(1), Constellation hereby requests Nuclear Regulatory Commission (NRC) approval of Revision 1 of the Reference 2 proposed alternative for Braidwood Generating Station (Braidwood), Units 1 and 2; Byron Generating Station (Byron),
Units 1 and 2; Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2; and R.E.
Ginna Nuclear Power Plant (Ginna). This proposed alternative requests to extend the inspection interval for the ASME Code, Section XI, Table IWB- 2500-1, Examination Categories B-B and B-D, and Table IWC-2500-1, Examination Category C-A steam generator examinations for the remainder of the currently licensed operating periods.
The proposed alternative is to extend the frequency of volumetric examinations of steam generator pressure retaining welds and full penetration welded nozzles for the remainder of the currently licensed operating periods for Braidwood, Units 1 and 2; Byron, Units 1 and 2; Calvert Cliffs, Units 1 and 2; and Ginna, and is provided in the Attachment 1 to this letter.
The proposed alternative includes a performance monitoring plan which will provide direct evidence to the presence and extend of any degradation, confirm the continued adequacy of the analysis, and provide a timely method to detect any novel or unexpected degradation.
Constellation requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that the alternative provides an acceptable level of quality and safety.
NRC review and approval of the proposed alternative is requested by March 15, 2024, for Braidwood Generating Station to support the Spring 2024 outage when the subject examinations are currently scheduled. NRC review and approval of the proposed alternative for the remaining stations is requested by August 1, 2024.
If you have any questions or require additional information, please contact Tom Loomis at (267) 533-5669.
Respectfully, David T. Gudger Senior Manager - Licensing and Regulatory Affairs Constellation Generation Company, LLC
 
Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles October 11, 2023 Page 3 :  10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2; Proposed Alternative I4R-23 for Byron Station, Units 1 and 2; Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2; and Proposed Alternative I6R-10 for Ginna, Revision 1 cc:    Regional Administrator - NRC Region I Regional Administrator - NRC Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant NRC Senior Resident Inspector - Ginna NRC Project Manager - Braidwood Station NRC Project Manager - Byron Station NRC Project Manager - Calvert Cliffs Nuclear Power Plant NRC Project Manager - Ginna S. Seaman, State of Maryland A. L. Peterson, NYSERDA
 
Attachment 1 Braidwood Station Byron Station Calvert Cliffs Ginna 10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna, Revision 1
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 1 of 98)
Proposed Alternative for Examinations of Pressure Retaining Welds and Full Penetration Welded Nozzles (Examination Categories B-B, B-D, and C-A)
In Accordance with 10 CFR 50.55a(z)(1) 1      ASME CODE COMPONENTS AFFECTED:
Code Class:                Class 1 and Class 2
 
== Description:==
Steam Generator (SG) Pressure Retaining Welds and Full Penetration Welded Nozzles Examination Categories: Category B-B, Pressure Retaining Welds in Vessels Other Than Reactor Vessels Category B-D, Full Penetration Welded Nozzles in Vessels Category C-A, Pressure Retaining Welds in Pressure Vessels Item Numbers:              B2.40 - Steam Generators (Primary Side), Tubesheet-to-Head Weld B3.130 - Steam Generators (Primary Side), Nozzle-to-Vessel Welds C1.10 - Shell Circumferential Welds C1.20 - Head Circumferential Welds C1.30 - Tubesheet-to-Shell Weld Braidwood Component IDs:
Unit  SG        Component ID    Item Number                    Description 1      A      1SG-05-SGC-01        B2.40                Primary Head - Tubesheet 1      A      1SG-05-SGC-05        C1.10          Upper Secondary Shell - Shell Cone 1      A      1SG-05-SGC-06        C1.10          Shell Cone - Steam Drum Lower Shell 1      A      1SG-05-SGC-08        C1.20      Steam Drum Upper Shell - Steam Drum Head 1      A      1SG-05-SGC-02        C1.30          Tubesheet - Lower Secondary Shell 1      B      1SG-06-SGC-01        B2.40                Primary Head - Tubesheet 1      B      1SG-06-SGC-05        C1.10          Upper Secondary Shell - Shell Cone 1      B      1SG-06-SGC-06        C1.10          Shell Cone - Steam Drum Lower Shell 1      B      1SG-06-SGC-08        C1.20      Steam Drum Upper Shell - Steam Drum Head 1      B      1SG-06-SGC-02        C1.30          Tubesheet - Lower Secondary Shell 1      C      1SG-07-SGC-01        B2.40                Primary Head - Tubesheet 1      C      1SG-07-SGC-05        C1.10          Upper Secondary Shell - Shell Cone 1      C      1SG-07-SGC-06        C1.10          Shell Cone - Steam Drum Lower Shell 1      C      1SG-07-SGC-08        C1.20      Steam Drum Upper Shell - Steam Drum Head 1      C      1SG-07-SGC-02        C1.30          Tubesheet - Lower Secondary Shell 1      D      1SG-08-SGC-01        B2.40                Primary Head - Tubesheet 1      D      1SG-08-SGC-05        C1.10          Upper Secondary Shell - Shell Cone 1      D      1SG-08-SGC-06        C1.10          Shell Cone - Steam Drum Lower Shell 1      D      1SG-08-SGC-08        C1.20      Steam Drum Upper Shell - Steam Drum Head 1      D      1SG-08-SGC-02        C1.30          Tubesheet - Lower Secondary Shell
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 2 of 98)
Unit    SG        Component ID      Item Number                        Description 2      A        2SG-01-SGC-01          B2.40                  Channel Head - Tubesheet 2      A        2SG-01-SGC-03          C1.10                  Stub Barrel - Lower Shell (A) 2      A        2SG-01-SGC-05          C1.10              Lower Shell (B) - Transition Cone 2      A        2SG-01-SGC-06          C1.10              Transition Cone - Upper Shell (A) 2      A        2SG-01-SGC-08          C1.20                Upper Shell (B) - Upper Head 2      A        2SG-01-SGC-02          C1.30                    Tubesheet - Stub Barrel 2      B        2SG-02-SGC-01          B2.40                  Channel Head - Tubesheet 2      B        2SG-02-SGC-03          C1.10                  Stub Barrel - Lower Shell (A) 2      B        2SG-02-SGC-05          C1.10              Lower Shell (B) - Transition Cone 2      B        2SG-02-SGC-06          C1.10              Transition Cone - Upper Shell (A) 2      B        2SG-02-SGC-08          C1.20                Upper Shell (B) - Upper Head 2      B        2SG-02-SGC-02          C1.30                    Tubesheet - Stub Barrel 2      C        2SG-03-SGC-01          B2.40                  Channel Head - Tubesheet 2      C        2SG-03-SGC-03          C1.10                  Stub Barrel - Lower Shell (A) 2      C        2SG-03-SGC-05          C1.10              Lower Shell (B) - Transition Cone 2      C        2SG-03-SGC-06          C1.10              Transition Cone - Upper Shell (A) 2      C        2SG-03-SGC-08          C1.20                Upper Shell (B) - Upper Head 2      C        2SG-03-SGC-02          C1.30                    Tubesheet - Stub Barrel 2      D        2SG-04-SGC-01          B2.40                  Channel Head - Tubesheet 2      D        2SG-04-SGC-03          C1.10                  Stub Barrel - Lower Shell (A) 2      D        2SG-04-SGC-05          C1.10              Lower Shell (B) - Transition Cone 2      D        2SG-04-SGC-06          C1.10              Transition Cone - Upper Shell (A) 2      D        2SG-04-SGC-08          C1.20                Upper Shell (B) - Upper Head 2      D        2SG-04-SGC-02          C1.30                    Tubesheet - Stub Barrel
      *Differences in number of components between Unit 1 and Unit 2 is due to the replacement of the Unit 1 Steam Generators during the 1998 outage.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 3 of 98)
Byron Component IDs:
Unit    SG        Component ID      Item Number                        Description 1      A      1RC-01-BA/SGW-01        B2.40                  Tubesheet - Primary Head 1      A      1RC-01-BA/SGW-05        C1.10                Secondary Shell - Shell Cone 1      A      1RC-01-BA/SGW-06        C1.10                  Steam Drum - Shell Cone 1      A      1RC-01-BA/SGW-08        C1.20              Steam Drum Head - Steam Drum 1      A      1RC-01-BA/SGW-02        C1.30                Secondary Shell - Tubesheet 1      B      1RC-01-BB/SGW-01        B2.40                  Tubesheet - Primary Head 1      B      1RC-01-BB/SGW-05        C1.10                Secondary Shell - Shell Cone 1      B      1RC-01-BB/SGW-06        C1.10                  Steam Drum - Shell Cone 1      B      1RC-01-BB/SGW-08        C1.20              Steam Drum Head - Steam Drum 1      B      1RC-01-BB/SGW-02        C1.30                Secondary Shell - Tubesheet 1      C      1RC-01-BC/SGW-01        B2.40                  Tubesheet - Primary Head 1      C      1RC-01-BC/SGW-05        C1.10                Secondary Shell - Shell Cone 1      C      1RC-01-BC/SGW-06        C1.10                  Steam Drum - Shell Cone 1      C      1RC-01-BC/SGW-08        C1.20              Steam Drum Head - Steam Drum 1      C      1RC-01-BC/SGW-02        C1.30                Secondary Shell - Tubesheet 1      D      1RC-01-BD/SGW-01        B2.40                  Tubesheet - Primary Head 1      D      1RC-01-BD/SGW-05        C1.10                Secondary Shell - Shell Cone 1      D      1RC-01-BD/SGW-06        C1.10                  Steam Drum - Shell Cone 1      D      1RC-01-BD/SGW-08        C1.20              Steam Drum Head - Steam Drum 1      D      1RC-01-BD/SGW-02        C1.30                Secondary Shell - Tubesheet 2      A      2RC-01-BA/SGC-01        B2.40                  Stub Barrel - Channel Head 2      A      2RC-01-BA/SGC-03        C1.10                Stub Barrel - Lower Barrel A 2      A      2RC-01-BA/SGC-05        C1.10              Lower Barrel B - Transition Cone 2      A      2RC-01-BA/SGC-06        C1.10              Transition Cone - Upper Barrel B 2      A      2RC-01-BA/SGC-08        C1.20                Upper Barrel B - Upper Head 2      A      2RC-01-BA/SGC-02        C1.30                        Stub Barrel Weld 2      B      2RC-01-BB/SGC-01        B2.40                  Stub Barrel - Channel Head 2      B      2RC-01-BB/SGC-03        C1.10                Stub Barrel - Lower Barrel A 2      B      2RC-01-BB/SGC-05        C1.10              Lower Barrel B - Transition Cone 2      B      2RC-01-BB/SGC-06        C1.10              Transition Cone - Upper Barrel B 2      B      2RC-01-BB/SGC-08        C1.20                Upper Barrel B - Upper Head 2      B      2RC-01-BB/SGC-02        C1.30                        Stub Barrel Weld 2      C      2RC-01-BC/SGC-01        B2.40                  Stub Barrel - Channel Head 2      C      2RC-01-BC/SGC-03        C1.10                Stub Barrel - Lower Barrel A 2      C      2RC-01-BC/SGC-05        C1.10              Lower Barrel B - Transition Cone 2      C      2RC-01-BC/SGC-06        C1.10              Transition Cone - Upper Barrel B 2      C      2RC-01-BC/SGC-08        C1.20                Upper Barrel B - Upper Head 2      C      2RC-01-BC/SGC-02        C1.30                        Stub Barrel Weld 2      D      2RC-01-BD/SGC-01        B2.40                  Stub Barrel - Channel Head 2      D      2RC-01-BD/SGC-03        C1.10                Stub Barrel - Lower Barrel A 2      D      2RC-01-BD/SGC-05        C1.10              Lower Barrel B - Transition Cone 2      D      2RC-01-BD/SGC-06        C1.10              Transition Cone - Upper Barrel B 2      D      2RC-01-BD/SGC-08        C1.20                Upper Barrel B - Upper Head 2      D      2RC-01-BD/SGC-02        C1.30                        Stub Barrel Weld
      *Differences in number of components between Unit 1 and Unit 2 is due to the replacement of the Unit 1 Steam Generators during the 1997 outage.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 4 of 98)
Calvert Cliffs Component IDs:
Unit    SG        Component ID        Item Number                        Description 1      11            11-W22            B2.40                  Tubesheet to Primary Head 1      11          SG W5            B3.130      11 Primary Outlet Nozzle to Primary Head (Hot Leg) 1      11          SG W6            B3.130            11A Primary Head to Cold Leg A Nozzle 1      11          SG W7            B3.130            11B Primary Head to Cold Leg B Nozzle 1      11            SG-11-4            C1.10            Upper Assembly to Transition Assembly 1      11            SG-11-3            C1.20                Upper Head to Upper Assembly 1      11        SG W65            C1.30                  Shell Can #1 to Tubesheet 1      12            12-W22            B2.40                  Tubesheet to Primary Head 1      12          SG W5          B3.130      12 Primary Outlet Nozzle to Primary Head (Hot Leg) 1      12          SG W6          B3.130            12A Primary Head to Cold Leg A Nozzle 1      12          SG W7          B3.130            12B Primary Head to Cold Leg B Nozzle 1      12            SG-12-4            C1.10            Upper Assembly to Transition Assembly 1      12            SG-12-3            C1.20                Upper Head to Upper Assembly 1      12        SG W65            C1.30                  Shell Can #1 to Tubesheet 2      21            21-W22            B2.40                  Tubesheet to Primary Head 2      21          SG W5          B3.130      21 Primary Outlet Nozzle to Primary Head (Hot Leg) 2      21          SG W6          B3.130            21A Primary Head to Cold Leg A Nozzle 2      21          SG W7          B3.130            21B Primary Head to Cold Leg B Nozzle 2      21            SG-21-4            C1.10            Upper Assembly to Transition Assembly 2      21            SG-21-3            C1.20                Upper Head to Upper Assembly 2      21        SG W65            C1.30                  Shell Can #1 to Tubesheet 2      22            22-W22            B2.40                  Tubesheet to Primary Head 2      22          SG W5          B3.130      22 Primary Outlet Nozzle to Primary Head (Hot Leg) 2      22          SG W6          B3.130            22A Primary Head to Cold Leg A Nozzle 2      22          SG W7          B3.130            22B Primary Head to Cold Leg B Nozzle 2      22            SG-22-4            C1.10            Upper Assembly to Transition Assembly 2      22            SG-22-3            C1.20                Upper Head to Upper Assembly 2      22        SG-22 -W65            C1.30                  Shell Can #1 to Tubesheet
* Partial replacements of the Unit 1 Steam Generators were performed during the 2002 outage. Partial replacements of the U2 Steam Generators were performed during the 2003 outage. The steam drum, which includes the Upper Assembly to Transition Assembly and Upper Head to Upper Assembly welds were retained and refurbished as part of the replacements. Therefore, these welds were retained. Examination records prior to the Steam Generator replacements could not be located; however, two inspections intervals of examinations have been performed since the Steam Generator replacements with no unacceptable indications identified.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 5 of 98)
Ginna Component IDs:
Unit    SG        Component ID      Item Number                        Description 1      A        LHTSW-AR            B2.40                Lower Head to Tubesheet Weld 1      A            LST-AR            C1.10              Lower Shell to Transition Circ Weld 1      A          TUS-AR            C1.10              Transition to Upper Shell Circ Weld 1      A          USH-AR            C1.20                  Upper Shell to Head Circ Weld 1      A          TSSW-AR            C1.30                    Tubesheet to Shell Weld 1      B        LHTSW-BR            B2.40                Lower Head to Tubesheet Weld 1      B            LST-BR            C1.10              Lower Shell to Transition Circ Weld 1      B          TUS-BR            C1.10              Transition to Upper Shell Circ Weld 1      B          USH-BR            C1.20                  Upper Shell to Head Circ Weld 1      B          TSSW-BR            C1.30                    Tubesheet to Shell Weld
        *Replacement of the Unit 1 Steam Generators were performed during the 1996 outage.
2        APPLICABLE CODE EDITION AND ADDENDA:
The following table identifies the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Section XI Code of Record for performing Inservice Inspection (ISI) activities at Braidwood, Byron, Calvert Cliffs, and Ginna:
Current Edition Plant            Interval                          Interval Start              Interval End and Addenda Braidwood Station, Units 1                                August 29, 2018 (Unit 1)    July 28, 2028 (Unit 1)
Fourth      2013 Edition and 2                                          November 5, 2018 (Unit 2) October 16, 2028 (Unit 2)
Byron Station, Units 1 and              2007 Edition with Fourth                          July 16, 2016              July 15, 2025 2                        the 2008 Addenda Calvert Cliffs Nuclear Fifth      2013 Edition          July 1, 2019              June 30, 2029 Power Plant, Units 1 and 2 R.E. Ginna Nuclear Power Sixth      2013 Edition        January 1, 2020          December 31, 2029 Plant The 2019 Edition of ASME Section XI, Table G-2110-1 will be utilized to extend the use of Figure G-2110-1, Reference Critical Stress Intensity Factor for Material, to material SA-533, Grade A, Class 2. (Note: The 2019 Edition of ASME Section XI is approved in 10 CFR 50.55a.).
3        APPLICABLE CODE REQUIREMENT:
The 2007 Edition with the 2008 Addenda and 2013 Edition of ASME Code, Section XI (ASME Section XI), IWB-2500(a), Table IWB-2500-1, Examination Categories B-B and B-D, and IWC-2500(a), Table IWC-2500-1, Examination Category C-A, require examination of the applicable Item Numbers:
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 6 of 98)
Item Number B2.40 - Volumetric examination of essentially 100% of the weld length. The examination may be limited to one vessel among the group of vessels performing a similar function during successive inspection intervals. The examination volume is shown in Figure IWB-2500-6.
Item Number B3.130 - Volumetric examination of all nozzles during each inspection interval. The examination volume is shown in Figures IWB-2500-7(a),
(b), (c) or (d).
Item Number C1.10 - Volumetric examination of essentially 100% of the weld length of the cylindrical-shell-to-conical shell-junction welds and shell (or head)-
to-flange welds during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-1.
Item Number C1.20 - Volumetric examination of essentially 100% of the weld length of the head-to-shell weld during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators),
the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-1.
Item Number C1.30 - Volumetric examination of essentially 100% of the weld length of the tubesheet-to-shell weld during each inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-2.
4    REASON FOR REQUEST:
The Electric Power Research Institute (EPRI) performed assessments in Reference [1]
of the basis for the ASME Section XI examination requirements specified for the above-listed ASME Section XI Examination Categories and Item Numbers for Steam Generator (SG) welds and components. The assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [1] report concluded that the current ASME Section XI inspection frequency of ten years can be increased significantly with no impact to plant safety. It is upon the basis of this conclusion that an alternate inspection frequency is being requested. The Reference [1] report was developed consistent with the recommendations provided in EPRls White Paper on PFM Reference [2].
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 7 of 98) 5    PROPOSED ALTERNATIVE AND BASIS FOR USE:
Constellation Energy Generation, LLC (Constellation) is requesting an inspection alternative to the examination requirements of ASME Section XI, IWB-2500(a), Table IWB-2500-1, Examination Categories B-B and B-D, and IWC-2500(a), Table IWC-2500-1, Examination Category C-A. The proposed alternative is to defer the inspection of these examination items from the current ASME Section XI requirement to the end of the currently approved Period of Extended Operation (PEO) for Braidwood Station (Braidwood), Units 1 and 2, Byron Station (Byron), Units 1 and 2, Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2, and Ginna, as summarized in the following table.
During review of Revision 0 of this proposed alternative [15], the NRC submitted requests for additional information [16] to support their review. Constellations response
[17] to RAI 2, RAI 3, RAI 4, RAI 5, RAI 6, RAI 7, RAI 8, and RAI 9 remains valid and is part of Revision 1 of the proposed alternative. Constellations response to RAI 1, related to performance monitoring, is no longer valid and is not part of Revision 1 of the proposed alternative.
Constellation provided supplemental information to support the RAI response [18, 19].
The supplemental information for RAI 2.a. of [18] remains valid and is part of Revision 1 of the proposed alternative. As Stated in RAI response [18]:
With regard to the response to RAI 2.a contained in Reference 6, the following clarification is provided. As stated in Reference 6, as used in the EPRI Report pre-service inspection (PSI) refers to the collective examinations required by ASME Code, Section III, during fabrication and any ASME Code, Section XI examinations performed prior to service. The EPRI Report also states These components did receive 100% coverage during PSI through a combination of the required Section III fabrication and Section XI PSI examinations. A signed N-1 or N-2 Certificate holders Data Report, as applicable, certifies that the steam generators were designed, fabricated, and examined in accordance with ASME Code, Section III, requirements. This includes meeting all requirements of NB-5000 or NC-5000, as applicable, for required examination of welds, acceptance standards, and qualification of personnel and provides assurance that complete coverage was achieved for the required fabrication examinations. Additionally, all the SG welds at the subject units also received ASME Code, Section XI, preservice examinations prior to initial service, which further contributed to through initial examinations and satisfies the assumption of the EPRI Report of 100% coverage during PSI.
All other information (and requests) in the RAI response [18, 19] are no longer valid and are not part of Revision 1 of the proposed alternative.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 8 of 98)
Revision 1 of the proposed alternative continues to rely on PFM technology as part of the technical basis. The NRC has stated during several public meetings that PFM is inherently risk-informed and consequently, any submittal that uses PFM technology as its basis will be reviewed as a risk-informed submittal. Therefore, Revision 1 of the proposed alternative includes a performance monitoring plan which will provide direct evidence to the presence and extent of any degradation, confirm the continued adequacy of the analysis, and provide a timely method to detect any novel or unexpected degradation.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 9 of 98)
Table 1. Summary of Inspection Deferrals in this Proposed Alternative End of Current Licensed Length of Time Until ASME      Item                                        Date of Last    Operating Period      Next Inspection for Station  Unit                                Description Category    No.                                        Inspection        (60 Years)          This Request (Years)
Braidwood  1    B-B      B2.40    Steam Generator (Primary Side),      10/8/2010        10/17/2046                36.0 TubesheetToHead Weld C-A      C1.10  Steam Generator (Secondary Side),    4/19/2018                                  28.5 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),      4/8/2015                                  31.5 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),    10/8/2010                                  36.0 TubesheettoShell Weld Braidwood  2    B-B      B2.40    Steam Generator (Primary Side),    10/26/2012        12/18/2047                35.2 TubesheetToHead Weld C-A      C1.10  Steam Generator (Secondary Side),    10/28/2021                                  26.2 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),    5/10/2014                                  33.6 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),    10/31/2018                                  29.2 TubesheettoShell Weld Byron    1    B-B      B2.40    Steam Generator (Primary Side),      3/2/2017        10/31/2044                27.7 TubesheetToHead Weld C-A      C1.10  Steam Generator (Secondary Side),      3/2023                                    21.6 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),    3/15/2020                                  24.6 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),    9/13/2018                                  26.2 TubesheettoShell Weld Byron    2    B-B      B2.40    Steam Generator (Primary Side),      4/13/2019          11/6/2046                27.6 TubesheetToHead Weld C-A      C1.10  Steam Generator (Secondary Side),    4/14/2013                                  33.6 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),    4/18/2013                                  33.6 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),    4/13/2019                                  27.6 TubesheettoShell Weld
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 10 of 98)
Table 1. Summary of Inspection Deferrals in this Proposed Alternative End of Current Licensed Length of Time Until ASME        Item                                          Date of Last        Operating Period      Next Inspection for Station    Unit                                    Description Category      No.                                          Inspection            (60 Years)          This Request (Years)
Calvert      1        B-B      B2.40    Steam Generator (Primary Side),      2/25/2014              7/31/2034                20.4 Cliffs                                      TubesheetToHead Weld B-D      B3.130    Steam Generator (Primary Side),      2/27/2018                                      16.4 NozzletoVessel Welds C-A      C1.10  Steam Generator (Secondary Side),      3/4/2018                                        16.4 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),      3/3/2018                                        16.4 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),      3/2/2018                                        16.4 TubesheettoShell Weld Calvert      2        B-B      B2.40    Steam Generator (Primary Side),      2/27/2013              8/13/2036                23.5 Cliffs                                      TubesheetToHead Weld B-D      B3.130    Steam Generator (Primary Side),      3/1/2013                                        23.5 NozzletoVessel Welds C-A      C1.10  Steam Generator (Secondary Side),      2/20/2011                                      25.5 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),      2/26/2019                                      17.5 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),      2/20/2017                                      19.5 TubesheettoShell Weld Ginna      1        B-B      B2.40    Steam Generator (Primary Side),      10/25/2015              9/18/2029                13.9 TubesheetToHead Weld C-A      C1.10  Steam Generator (Secondary Side),      5/1/2017                                        12.4 Shell Circumferential Welds C-A      C1.20  Steam Generator (Secondary Side),      10/31/2012                                      16.9 Head Circumferential Welds C-A      C1.30  Steam Generator (Secondary Side),      10/25/2015                                      13.9 TubesheettoShell Weld Notes:
1 - does not account for performance monitoring examinations that will be performed during the requested deferral period.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 11 of 98)
As indicated in Table 1, the proposed alternative results in a maximum effective operating period of 36 years from the last inspection for Item Numbers B2.40 and C1.30 for Braidwood Unit 1 included in this Proposed Alternative. As summarized in Attachment 1, the EPRI report demonstrates that after preservice inspection (PSI) and two 10-year intervals, the failure probabilities (for both rupture and leakage) are significantly below the acceptance criteria of 1x10-6 failures per year for time periods exceeding 36 years of operation.
The key aspects in the technical basis for this request are summarized below. The applicability of the technical basis for Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna, is shown in Appendix A. It is noted that the inputs for transients and cycle counting remain valid for this Revision 1 submittal.
Degradation Mechanism Evaluation An evaluation of degradation mechanisms that could potentially impact the reliability of the SG pressure retaining welds and full penetration welded nozzles was performed in Reference [1]. Evaluated mechanisms included stress corrosion cracking (SCC),
environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC),
pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC),
general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG pressure retaining welds and full penetration welded nozzles.
Stress Analysis Finite element analysis (FEA) was performed in Reference [1] to determine the stresses in the SG pressure retaining welds and full penetration welded nozzles. The analysis was performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties.
As stated in Section 9.0 of Reference [1], the operating transients and associated cycles evaluated are typical of historical PWR operations, although they can vary depending on the design type as well as within each design type. In the fracture mechanics evaluations performed in Reference [1], the transients that contribute the most to crack growth are the heatup and cooldown events. A total of 300 such events were evaluated during a 60-year plant life. Reference [1] also used a conservative rate of 200°F per hour which would exceed the limits allowed by the Braidwood, Byron, Calvert Cliffs, and Ginna technical specifications. The sensitivity study performed in Section 8.0 of Reference [1]
shows that, for the realistic ISI scenario of PSI followed by ISI at 20, 40, and 60 years, a stress multiplier of 1.9 can accommodate small variations in the pressures and temperatures of the evaluated transients. Due to the small variations from the values
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 12 of 98) from Reference [1], the conservative definition of the heatup and cooldown transient, and the low number of transient cycles predicted in Appendix A over the life of the plant vs. the number of cycles evaluated in Reference [1], it is determined that the stress analysis performed in Reference [1] is applicable to Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna.
The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna, are shown in Appendix A and confirms that all plant-specific requirements are met. Therefore, the evaluation results and conclusions of Reference [1] are applicable to Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna.
Flaw Tolerance Evaluation Flaw tolerance evaluations were performed in Reference [1] consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after the PSI followed by two 10-year intervals, no other inspections are required for up to 60 years of plant operation to meet the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year. For the specific case of Braidwood and Byron, Unit 1, PSI has been followed by the performance of two 10-year interval inspections. For Braidwood and Byron, Unit 2, PSI has been followed by the performance of three 10-year interval inspections. For Calvert Cliffs, Units 1 and 2, PSI has been followed by the performance of two 10-year interval inspections for welds W5, W6, W7, W22, and W65. For Calvert Cliffs, Units 1 and 2, PSI has been followed by the performance of four 10-year interval inspections for welds W3 and W4, but records were only found for the Third and Fourth Interval examinations. For Ginna, PSI has been followed by the performance of two 10-year interval inspections.
The most limiting case requested within this proposed alternative, analyzed by the EPRI report, is the scenario of PSI followed by two 10-year intervals, followed by a 30-year inspection (PSI + 10 + 20 + 50). The EPRI report does not contain Probability of Failure results for the limiting case PSI + 10 + 20 + 50, but the PSI + 20 + 40 + 60 case has similar Probability of Failure results, as shown in Table 8-10 of the EPRI report. The Probability of Failure results would be larger for the limiting PSI + 10 + 20 + 50 case than the PSI + 20 + 40 + 60 case, but they would be of the same magnitude. The Probability of Failure results in Table 8-9 for the PSI only case and Table 8-10 for the PSI + 20 + 40 + 60 case show a large margin from the criterion of 1x10-6 per year. The Probability of Failures listed in Table 8-10 only address the base case scenario and it does not address the uncertainties that are addressed in the sensitivity studies in Section 8.3.4.3 of the EPRI report. Therefore, results of the sensitivity studies should be applied to the limiting PSI + 10 + 20 + 50 case requested in this proposed alternative.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 13 of 98)
A sensitivity study was performed on the effects of various key parameters used in the EPRI report. For the sensitivity study, the mean fracture toughness was changed to 200 ksiin. with a conservative standard deviation of 30 ksiin. and the stress multiplier was changed to 1.9. The limiting case for the sensitivity study, that is applicable to this relief request, was found to be the primary tubesheet-to-shell weld with a probability of rupture of 7.33x10-7, for the PSI + 20 + 40 + 60 case, which is below the 1x10-6 per year acceptance criteria. All other locations reviewed by the sensitivity study, applicable to this relief request, were below the 1x10-6 per year acceptance criteria.
The sensitivity study results in the EPRI report are assuming 80 years of operation, but this proposed alternative only requests extension of the ISI examinations for no more than 60 years of operation. If the PFM analysis was performed for only 60 years of operation, the probability of failure values would be lower than the determined values.
The DFM evaluations confirm the PFM results by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to the ASME Section XI acceptance standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Section XI allowable fracture toughness.
Based on the information provided, the proposed alternative of PSI + 10 + 20 + 50 for the requested Steam Generator welds for the plants request, would result in a Probability of Failure per year that is reasonably below the criterion of 1x10-6 per year.
Inspection History Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna, operating experience (including examinations performed to date, examination findings, inspection coverage, and Request for Alternatives) are presented in Appendix B. As shown in Appendix B, Braidwood and Calvert Cliffs recorded examinations with limited coverage for the applicable components. The lowest recorded coverage for Braidwood was 88.4% for the Item Number C1.30 (tubesheet-to-shell welds), and for Calvert Cliffs the lowest coverage was 73.8% for the Item Number B3.130 (nozzle-to-vessel welds).
Based on the sensitivity study results presented in Table 8-33 of Reference [1], these coverage values meet the Probability of Rupture acceptance criterion for all locations when the life of the component reaches 60 years. Also, as shown in Appendix B, Byron was the only plant to identify recordable indications that exceeded the ASME Section XI acceptance standards. Welds 2RC-01-BB/SGC-03, 2RC-01-BC/SGC-05, and 2RC      BD/SGC-06 were found to have embedded indications during the PSI examinations exceeding the ASME Section XI acceptance standards. Relief to accept these flaws was approved through an NRC safety evaluation on 10/29/1986. Successive examinations were performed for each weld and the embedded indications were either found to have no change in sizing or were found to be non-recordable due to improved inspection methods during later ISI Intervals. Weld 2RC-01-BA/SGC-02 was also found to have 4
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 14 of 98) embedded indications during the 1st ISI Interval examination. Successive examinations were performed and the embedded indications were either found to be non-relevant or did not change in size during the latest 4th ISI Interval examination.
Industry inspection history for the applicable components (as obtained from the industry survey summarized in Reference [1]) is presented in Appendix C. The results of the survey in Reference [1] indicate that these components are very flaw tolerant.
Performance Monitoring Requirements In an April 27, 2023, public meeting (ML23114A034), the NRC stated that probabilistic fracture mechanics (PFM) is inherently risk-informed and consequently, any submittal that uses PFM technology as its basis, will be reviewed as a risk-informed submittal.
Therefore, performance monitoring must be addressed in any submittals relying on PFM.
An acceptable performance monitoring approach must provide:
* Direct evidence of the presence and/or extent of degradation;
* Validation and confirmation of continued adequacy of analysis, and;
* A timely method to detect novel or unexpected degradation.
In the same April public meeting, the NRC stated that a performance monitoring sample of 25% would yield acceptable results, under certain assumptions, for the US fleet of 61 PWRs using a binomial distribution. It was noted that a smaller performance monitoring sample may be acceptable using a Monte Carlo analysis method. This same 25%
performance monitoring sample was reiterated during the June 15, 2023, industry technical exchange meetings (ML23158A180). This performance monitoring sample would result in a 75% examination reduction compared to the current ASME Section XI examination requirements. Since this proposed alternative relies on PFM as part of the technical basis, Constellations proposed performance monitoring plan is provided below.
Note, this proposed performance monitoring plan covers items number included as part of this proposed alternative (B2.40, B3.130, C1.10, C1.20, and C1.30) and includes item number (C2.21 and C2.22) which are the subject of a separate proposed alternative submitted by Constellation.
Current Inspection Intervals for Braidwood, Byron, Calvert Cliffs, and Ginna The dates for the current inspection intervals are defined in Section 2 of the proposed alternative. To satisfy the ASME Section XI requirements applicable to the components covered by the proposed alternatives, the seven Constellation PWR units are required to perform a combined 67 examinations of steam generator welds and components each inspection interval. The table below identifies the number of required examinations, by
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 15 of 98)
Unit, Examination Category, and Item Number. The Comp. column identifies how many examinations have been completed for each Unit during the current ISI Interval. As shown in the table below, the seven Constellation PWR units have completed a combined 22 steam generator examinations (~33% of the required 67 examinations) during the current ISI Intervals. These examinations were performed starting in 2017, with the latest examinations performed during the Spring 2023 Byron Unit 1 outage (B1R25). Appendix B of the proposed alternatives provide the plant specific inspection history for each weld or component.
The examinations completed during the current, and previous, inspection intervals serve as sufficient performance monitoring to support the proposed alternative for the remainder of the current interval for the Constellation PWR units. The examinations performed during the current inspection intervals provide direct evidence of the absence of any service induced degradation, validate the adequacy of the PFM analysis, and confirm the absence of any novel or unexpected degradation. Therefore, no additional examinations will be performed during the current inspection intervals for any of the Constellation PWR units.
Ginna - Successive Inspection Intervals Ginnas current 6th ISI interval began on January 1, 2020, and will effectively end on September 18, 2029, with the expiration of the current operating license. No examinations are proposed to be performed at Ginna during the current 6th ISI interval for the components covered by this proposed alternative which will result in a maximum of 16.9 years of elapsed time between the completion of the last ASME XI examination and the end of the current operating license. Ginna most recently completed all ASME XI examination requirements for the components covered by the proposed alternatives during the 5th ISI interval which started on January 1, 2010, and ended on December 31, 2019. At this time no examinations have been completed during the 6th ISI Interval for the applicable components. Appendix B of the proposed alternatives provide the
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 16 of 98) complete examination history for each component at Ginna. The examinations completed during previous ISI intervals, along with the sample of examinations performed by the Constellation PWR units during the current inspection intervals, provide sufficient performance monitoring to support the proposed alternative for the remainder of the current interval for Ginna.
If Ginna pursues subsequent license renewal to operate beyond the end of the current operating license, the ASME Section XI examination requirements would again become applicable at that time. If Ginna desires to use a similar proposed alternative in the second period of extended operation, it is required to be addressed in the license renewal application or a separate proposed alternative submitted in accordance with 10 CFR 50.55a(z), either of which will require separate NRC review and approval.
Calvert Cliffs - Successive Inspection Intervals Calvert Cliffs current 5th ISI interval began on July 1, 2019, and is scheduled to conclude on June 30, 2029. During the current interval, Calvert Cliffs Unit 1 has performed 6 steam generator examinations with no unacceptable indications identified.
These examinations include 3 Class 1 nozzle-to-vessel welds (Category B-D), 1 Class 2 pressure retaining welds in pressure vessels (Category C-A), and 1 Class 2 nozzle to vessel weld and 1 inside radius section (Category C-B), as shown in the Table above.
For Calvert Cliffs, the examinations performed on Unit 1 during the current ISI interval, examinations performed during previous ISI intervals, and the sample of examinations performed by the Constellation PWR units during the current ISI intervals, provides sufficient performance monitoring to support the proposed alternative for the remainder of the current interval for Calvert Cliffs Unit 1 and 2.
During the successive 6th ISI interval, which is currently scheduled to begin on July 1, 2029, Calvert Cliffs will perform a sample of 25% of the ASME Section XI required examinations as a performance monitoring plan. Given that Unit 1 has performed examinations during the 5th ISI interval and that the Unit 2 operating license expires later, the performance monitoring examinations will be performed at Calvert Cliffs Unit 2.
The performance monitoring sample size will be prorated to account for the fact that the current operating licenses expire midway through the 6th ISI interval. The current operating license expires on July 31, 2034, for Unit 1 and August 13, 2036, for Unit 2.
Calvert Cliffs Unit 1 and Unit 2 will operate approximately 5 years and 7 years, respectively, in the 6th ISI interval prior to the expiration of the current operating license.
Therefore, as a performance monitoring plan, Calvert Cliffs Unit 2 will perform 5 steam generator examinations during the 6th ISI interval as determined below:
14 exams per unit per int.
* 2 units
* 0.25 PM sample
* 12 yr. / 20 yr. proration = 5 exams
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 17 of 98)
The examinations to be performed as part of the performance monitoring sample will be selected during the 6th interval code of record update and scheduled in the ISI database. At least one component in each of the applicable Examination Categories will be selected for examination in the 6th ISI interval to ensure a representative sample of examinations are performed. The performance monitoring sample will be distributed between the first and second periods to ensure a continuous stream of data and allow for timely identification of any service induced degradation or the emergence of any novel degradation mechanisms. The components eligible for selection as part of the performance monitoring plan are shown in Section 1 of the proposed alternatives.
In the unlikely event that any new unacceptable indications (i.e., new indications exceeding the acceptance standards of IWB-3500 or IWC-3500, as applicable, that are accepted by Repair/Replacement Activity or analytical evaluation) are identified during the performance monitoring examinations at Calvert Cliffs Unit 2, the indications will be evaluated as required by ASME Section XI, and Constellation corrective action program.
The additional examination and successive inspection requirements of ASME Code, Section XI, also apply. The number of additional examinations, to be performed during the current refueling outage, shall be the number of performance monitoring examinations included in the inspection item number that were scheduled to be performed during the current inspection period. If additional unacceptable indications are identified, the examinations shall be further expanded to include all remaining welds/components in the inspection item number. Any new unacceptable indications identified as part of the performance monitoring examinations at Calvert Cliffs Unit 2 will result in both Units reverting to ASME Section XI examination requirements for the remainder of the 6th ISI interval for components covered by the proposed alternatives.
ASME Section XI examinations shall resume no later than the first or second refueling outage following discovery of the initial unacceptable indication. For Calvert Cliffs Unit 1, the same weld/component with the initial unacceptable indication discovered in Unit 2, shall be included for examination within the first or second refueling outage following discovery.
If Calvert Cliffs pursues subsequent license renewal to operate beyond the end of the current operating licenses, the ASME Section XI examination requirements would again become applicable at that time. If Calvert Cliffs desires to use a similar proposed alternative in the second period of extended operation, it is required to be addressed in the license renewal application or a separate proposed alternative submitted in accordance with 10 CFR 50.55a(z), either of which will require separate NRC review and approval.
Braidwood and Byron - Successive Inspection Intervals Given the similarities in design, materials, construction methods, service conditions, and operating strategies between Braidwood and Byron, a common performance monitoring
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 18 of 98) plan is being proposed for the current and successive inspection intervals. As described previously, a sufficient number of examinations have been performed at Braidwood, Byron, and across the Constellation PWR fleet to support the proposed alternative for the remainder of the current 4th ISI intervals. The proposed performance monitoring plan, applicable to the 5th and 6th ISI intervals at Braidwood and Byron, is for 1 out of the 4 units to complete all required ASME Section XI steam generator examinations each interval. This represents a 25% performance monitoring sample since one (of four required) steam generator worth of examinations will be performed during both the 5th and 6th ISI intervals. The units selected for performance monitoring will distribute the examinations across the periods as required by Table IWB-2411-1 or Table IWC-2411-1, as applicable. This will ensure a continuous stream of data and allow for timely identification of any service induced degradation or the emergence of any novel degradation mechanisms.
Given the limited number of examinations performed during the current ISI interval, Braidwood Unit 1 or Unit 2 will be selected for performance monitoring examinations during the 5th ISI interval. The 5th interval is currently scheduled to begin in July and October of 2028 for Braidwood Unit 1 and 2, respectively. Byron Unit 1 or Byron Unit 2 will be selected for performance monitoring examinations during the 6th ISI interval. The specific unit and components selected for examination will be determined during the ISI interval code of record updates and scheduled in the ISI database. The components eligible for selection as part of the performance monitoring plan are shown in Section 1 of the proposed alternatives.
In the unlikely event that any new unacceptable indications (i.e., new indications exceeding the acceptance standards of IWB-3500 or IWC-3500, as applicable, that are accepted by Repair/Replacement Activity or analytical evaluation) are identified during the performance monitoring examinations, the indications will be evaluated as required by ASME Section XI, and Constellation corrective action program. The additional examination and successive inspection requirements of ASME Code, Section XI, also apply. The number of additional examinations, to be performed during the current refueling outage, shall be the number of performance monitoring examinations included in the inspection item number that were scheduled to be performed during the current inspection period. If additional unacceptable indications are identified, the examinations shall be further expanded to include all remaining welds/components in the inspection item number. Any new unacceptable indications identified as part of the performance monitoring examinations at Braidwood or Byron will result in all four Units reverting to ASME Section XI examination requirements for the remainder of the interval during which the unacceptable indications are identified. ASME Section XI examinations shall resume no later than the first or second refueling outage following discovery of the initial unacceptable indication. The same weld/component with the initial unacceptable indication shall be included for examination at the remaining Braidwood and Byron Units within the first or second refueling outage following discovery. If the unacceptable
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 19 of 98) indication is identified during the 5th inspection interval, and no further unacceptable indications are identified during the resumption of ASME XI examinations, then the proposed alternative may continue to be applied for the 6th inspection interval for all Braidwood and Byron Units.
If Braidwood or Byron pursue subsequent license renewal to operate beyond the end of the current operating licenses, the ASME Section XI examination requirements would again become applicable at that time. If Braidwood or Byron desire to use a similar proposed alternative in the second period of extended operation, it is required to be addressed in the license renewal application or a separate proposed alternative will be submitted in accordance with 10 CFR 50.55a(z), either of which will require separate NRC review and approval.
Performance Monitoring Summary The proposed performance monitoring plan for the Constellation PWR fleet will provide:
* Direct evidence of the presence and/or extent of degradation;
* Validation and confirmation of continued adequacy of analysis, and;
* A timely method to detect novel or unexpected degradation.
A graphical representation of the performance monitoring plan is included in the Figure below. Distributing the examinations as required by IWB-2411-1, IWC-2411-1, or as described in the proposed alternative, will ensure that examinations are performed in each inspection period shown in the Figure below for the selected units at Braidwood, Byron, and Calvert Cliffs. Distributing the performance monitoring examinations will ensure a continuous stream of examination data for the duration of the proposed alternative.
In addition to the direct evidence provided by the performance monitoring examinations at Calvert Cliffs, Braidwood, and Byron, examination of steam generator welds and components is expected to continue to be performed by other units across the domestic and international PWR fleet. Continued examination of these steam generator components across the industry will provide additional opportunities to detect known degradation mechanisms and will also provide the opportunity to detect any new or unexpected degradation mechanisms that may occur in the future in the subject components. If a new degradation mechanism is identified during continued industry examinations, Constellation will follow the industry guidance to address the new degradation mechanism.
The absence of any new unacceptable indications in the Calvert Cliffs, Braidwood, and Byron performance monitoring examinations and the absence of any unexpected degradation across the operating fleet provides validation that the assumptions and
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 20 of 98) methods of the PFM Model used in the EPRI Reports are adequate to predict the future behavior of the subject welds and components. The strong technical basis provided by the results of the PFM Model and EPRI Report, along with the implementation of the proposed performance monitoring plan, including scope expansion criteria, will provide additional assurance that the steam generator components across the Constellation PWR fleet can operate safely for the remainder of the current operating licenses and will continue to provide an acceptable level of quality and safety.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 21 of 98)
Graphical Representation of Constellation Proposed Performance Monitoring Plan Plant        Year 2016 2017    2018 2019    2020    2021  2022  2023    2024    2025  2026    2027    2028  2029    2030    2031  2032  2033    2034  2035 2036    2037 2038 2039    2040    2041  2042    2043    2044    2045    2046    2047 4th Interval                                                              5th Interval                                                  6th Interval
* Braidwood 1 Period 1                Period 2                  Period 3              Period 1                  Period 2                Period 3          Period 1                  Period 2            Period 3
* 4th Interval                                                              5th Interval                                                      6th Interval
* Braidwood 2 Period 1                  Period 2                Period 3              Period 1                    Period 2              Period 3          Period 1                    Period 2              Period 3
* 4th Interval                                                        5th Interval                                                      6th Interval
* Byron 1 Period 1            Period 2            Period 3              Period 1                  Period 2                  Period 3            Period 1              Period 2                Period 3
* 4th Interval                                                        5th Interval                                                              6th Interval
* Byron 2 Period 1            Period 2            Period 3              Period 1                  Period 2                  Period 3            Period 1              Period 2                        Period 3
* 5th Interval                                            6th Interval
* Calvert Cliffs 1 Period 1                Period 2                  Period 3              Period 1          Period 2
* 5th Interval                                                    6th Interval
* Calvert Cliffs 2 Period 1                Period 2                  Period 3              Period 1                  Period 2
* 6th Interval
* Ginna Period 1                  Period 2                Period 3
* Note: The end dates for the current interval and the start and end dates for successive intervals are estimates based on the current interval. Interval start and end dates may be adjusted as allowed by ASME Section XI. Successive interval period start and end dates are approximate based on current interval scheduling.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 22 of 98)
Conclusion It is concluded that the SG pressure retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis in Reference [1] demonstrate that, after PSI followed by two 10-year intervals, no other inspection is required until 60 years to meet the NRC safety goal of 10-6 failures per reactor year. Plant-specific applicability of the technical basis to Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna are demonstrated in Appendix A. While the technical bases demonstrate longer inspection frequencies are possible, Constellation considers that deferral of these inspections until the end of the currently approved PEO, as shown in Table 1, with the proposed performance monitoring plan, provides an acceptable level of quality and safety in lieu of the current ASME Examination Categories B-B, B-D, and C-A, Item Numbers B2.40, B3.130, C1.10, C1.20, and C1.30 volumetric examination inspection frequency.
Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. As shown in Appendix B, to date, Constellation has performed over 110 inspections of SG pressure retaining welds and full penetration welded nozzles at Braidwood, Units 1 and 2, Byron, Units 1 and 2, Calvert Cliffs, Units 1 and 2, and Ginna. The flaws exceeding the ASME Section XI acceptance standards were fabrication flaws detected during PSI and are documented in Appendix B. Successive examinations have been performed for all recordable indications and there has been no increase in flaw sizing or the flaws are no longer recordable due to improved inspection methods. The lowest inspection coverage listed in Appendix B was 73.8%. Section 8.3.5 of Reference [1] discusses limited coverage and determines that the conclusions of the report are applicable to components with limited coverage greater than 50%. In addition, it is important to note all other inspection activities, including the system leakage test (Examination Categories B-P and C-H) conducted each outage or each ISI period, as applicable, will continue to be performed providing further assurance of safety.
The proposed performance monitoring plan will gather continuous data for the Constellation PWR fleet for the duration of the proposed alternative which will provide direct evidence of the presence and/or extent of degradation, validation of the continued adequacy of the analysis, and a timely method to detect novel or unexpected degradation. Therefore, Constellation requests that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).
6    DURATION OF PROPOSED ALTERNATIVE:
The proposed alternative is requested for Braidwood, Byron, Calvert Cliffs, and Ginna for the remainder of their currently approved operating license, currently scheduled to end on October 17, 2046 (Braidwood, Unit 1), December 18, 2047 (Braidwood, Unit 2),
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 23 of 98)
October 31, 2044 (Byron, Unit 1), November 6, 2046 (Byron, Unit 2), July 31, 2034 (Calvert Cliffs, Unit 1), August 13, 2036 (Calvert Cliffs, Unit 2), and September 18, 2029 (Ginna), as summarized in Table 1.
7    PRECEDENT:
To-date, there has been one previous submittal that request relief from the ASME Examination Categories B-B, B-D, and C-A, Item Numbers B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30 volumetric examinations on the basis of the Reference [1]
technical basis:
* Letter from M. Sartain (Dominion Energy) to U.S. Nuclear Regulatory Commission, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, July 15, 2020, ADAMS Accession No. ML20198M682.
* Letter from G. T. Bischof (Dominion Energy) to U.S. Nuclear Regulatory Commission, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Response to Request for Additional Information for Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, March 19, 2021, ADAMS Accession No. ML21081A136.
The following is a list of approved Request for Alternatives related to inspections of SG welds and components:
* Letter from N. L. Salgado (NRC) to B. C. Hanson (Exelon Generation Company, LLC), Braidwood Station, Units 1 and 2 - Relief from the Requirements of the ASME Code (EPID L-2019-LLR-0081), May 14, 2020, ADAMS Accession No. ML20133K093.
* Letter from J. G. Danna (NRC) to D. P. Rhoades (Exelon Generation Company, LLC), Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code Concerning Volumetric or Surface Examination Coverage for the Subject Welds (EPID L-2020-LLR-0089), January 15, 2021, ADAMS Accession No. ML20356A121.
* Letter from L. N. Olshan (NRC) to Dennis L Farrar (Commonwealth Edison Company), Approval of Byron 2 Preservice Inspection Program, dated October 29, 1986, 8611070024.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 24 of 98)
* Letter from Jamie M. Heisserer (NRC) to Jamie M. Coleman (Southern Nuclear Operating Co., Inc.), Joseph M. Farley Nuclear Plant, Units 1 and 2 - Proposed Inservice Inspection Alternative FNP-ISI-ALT-05-05, Version 1.0, To the Requirements of the ASME Code (EPID L-2022-LLR-0068), dated August 30, 2023, ADAMS Accession No. ML23164A120.
* Letter from Dave J Wrona (NRC) to Shawn Gibby (Duke Energy), Catawba Nuclear Station, Unit Nos. 1 and 2; Shearon Harris Nuclear Power Plant, Unit No.
1: McGuire Nuclear Station, Unit Nos. 1 and 2; Oconee Nuclear Station, Unit Nos. 1, 2, and 3; and H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of Alternative to Steam Generator Welds (EPID L-2023LLR-0003), dated September 25, 2023, ADAMS Accession No. ML23256A088.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 25 of 98) 8    ACRONYMS:
ASME      American Society of Mechanical Engineers B&W      Babcock and Wilcox BWR      Boiling Water Reactor BWRVIP    Boiling Water Reactor Vessel and Internals Program CE        Combustion Engineering CFR      Code of Federal Regulations DFM      Deterministic fracture mechanics EAF      Environmentally assisted fatigue EPRI      Electric Power Research Institute FAC      Flow accelerated corrosion FEA      Finite element analysis FW        Feedwater ISI      Inservice Inspection MIC      Microbiologically influenced corrosion MS        Main Steam NPS      Nominal pipe size NRC      Nuclear Regulatory Commission NSSS      Nuclear steam supply system PFM      Probabilistic fracture mechanics PWR      Pressurized Water Reactor SCC      Stress corrosion cracking SG        Steam Generator
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 26 of 98) 9   
 
==REFERENCES:==
: 1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906, ADAMS Accession No. ML20225A141.
: 2. N. Palm (EPRI), BWR Vessel & Internals Project (BWRVIP) Memo No. 2019- 016, White Paper on Suggested Content for PFM Submittals to the NRC, February 27, 2019, ADAMS Accession No. ML19241A545.
: 3. American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)
Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.
: 4. B. A. Bishop, C. Boggess, N. Palm, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Rev. 3, October 2011.
: 5. U.S. NRC, Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011, ADAMS Accession No. ML111600303.
: 6. BWRVIP-108: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to- Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557.
: 7. U.S. NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108), December 19, 2007, ADAMS Accession No. ML073600374.
: 8. BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.
: 9. U.S. NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241), April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.
: 10. Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, ASME Code Section XI, Division 1, Approval Date: February 20, 2004.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 27 of 98)
: 11. U.S. NRC Regulatory Guide 1.147, Revision 19, lnservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, March 2020.
: 12. American Society of Mechanical Engineers, Section XI Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with 2008 Addenda, 2013 Edition, and 2019 Edition.
: 13. U.S. NRC, Vogtle Electric Generating Plant, Units 1 and 2 - Audit Report for the Promise Version 1.0 Probabilistic Fracture Mechanics Software Used in Relief Request VEGP-ISI-ALT-04-04 (EPID L-2019-LLR-0109), December 10, 2020, ADAMS Accession No. ML20258A002.
: 14. U.S. NRC Regulatory Federal Register Volume 86, Issue 57, March 26, 2021.
: 15. Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles, dated December 14, 2021, ADAMS Accession No. ML21348A078.
: 16. Email from J. Wiebe (U.S. Nuclear Regulatory Commission) to T. Loomis (Constellation Energy Generation, LLC), Draft RAIs for Requests for Alternative I4R-17, I4R-23, ISI-05-018, I6R-10 (EPID Nos.: L-2021-LLR-091, L-2021-LLR-092, L-2021-LLR-093, L-2021-LLR-094), dated May 6, 2022, ADAMS Accession No. ML22129A013
: 17. Letter from D. Gudger (Constellation Energy Generation, LLC) to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information -
Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles, dated June 17, 2022, ADAMS Accession No. ML22168A005.
: 18. Letter from D. Reddick (Constellation Energy Generation, LLC) to U.S. Nuclear Regulatory Commission, Supplemental Information - Proposed Alternatives Related to the Steam Generators and Request for Forward Fit Analysis, dated September 20, 2022, ADAMS Accession No. ML22263A440.
: 19. Letter from D. Reddick (Constellation Energy Generation, LLC) to U.S. Nuclear Regulatory Commission, Supplemental Information - Proposed Alternatives Related to the Steam Generators, dated February 21, 2023, ADAMS Accession No. ML23052A068.
: 20. Letter from D. Reddick (Constellation Energy Generation, LLC) to U.S. Nuclear Regulatory Commission, Withdrawal - Proposed Alternatives Related to the Steam Generators, dated June 27, 2023, ADAMS Accession No. ML23178A169.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 28 of 98)
APPENDIX A PLANT SPECIFIC APPLICABILITY
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 29 of 98)
Plant-Specific Applicability for Braidwood Section 9 of Reference [A1] provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Braidwood, Units 1 and 2, is provided in Tables A1 through A4.
Tables A1 through A4 indicate that all plant-specific requirements are met for Braidwood, Units 1 and 2. Therefore, the results and conclusions of the EPRI report are applicable to Braidwood, Units 1 and 2.
Table A1. Applicability of Reference [A1] Representative Analyses to Braidwood, Units 1 and 2 Items Number B2.40 (SG Primary Side Shell Welds)
Requirement from              Applicability to              Applicability to Category Reference [A1]            Braidwood, Unit 1            Braidwood, Unit 2 General      The Loss of Power            Braidwood Station, Units 1 and 2 has not experienced Requirements    transient (involving        a loss of power transient resulting in unheated unheated auxiliary          auxiliary feedwater being introduced into a hot SG feedwater being              that has been boiled dry following blackout, resulting introduced into a hot        in thermal shock of portion of the vessel [A8].
SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 30 of 98)
The materials of the      The Braidwood          The Braidwood Station, Unit SG shell must be low      Station, Unit 1, SG    2 SG channel head is alloy ferritic steels    vessel primary head    fabricated of SA-216, Grade which conform to the      and tubesheet are      WCC material, and the SG requirements of ASME      fabricated of SA-508,  tube plate is fabricated from Code, Section XI,        Class 3 material. This SA-508, Class 2a material.
Appendix G,              material conforms to  Material SA-508, Class 2a Paragraph G- 2110.        the requirements of    conforms to the ASME Code, Section    requirements of ASME XI, Appendix G,        Code, Section XI, Appendix Paragraph G-2110.      G, Paragraph G- 2110.
SA-216 Gr. WCC material is not specifically listed in ASME Code, Section XI, Appendix G. However, SA-216 Grade WCC material has similar toughness and chemical composition to SA-533, Grade B, Class 1, SA-508-1, SA-508-2, and SA-508-3, it has a specified minimum yield strength at room temperature of 40 ksi (which is less than 50 ksi),
and the RTNDT values for the Braidwood Unit 2 steam generators channel head material was required to be 60°F or less (so the RTNDT of 60°F used in the EPRI report is bounding). Also, thermal transient stresses in SA-216 Grade WCC material are lower compared to the SA-533 Grade B Class 1 material used in the FEA in the EPRI report. Therefore, by comparison, SA-216 Gr.
WCC material is consistent with the requirements of ASME Code, Section XI, Appendix G and satisfies the requirements for application of the EPRI report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 31 of 98)
Requirement from            Applicability to              Applicability to Category Reference [A1]            Braidwood, Unit 1            Braidwood, Unit 2 Specific    The weld                  The Braidwood              The Braidwood Station, Unit Requirements    configurations must        Station, Unit 1 weld      2 weld configuration is conform to those          configuration is          shown in Figure A1 and shown in Figures 1-1      shown in Figure A1        conforms to Figure 1-2 of and Figure 1-2 of          and conforms to            Reference [A1].
Reference [A1].            Figure 1-2 of Reference [A1].
The SG vessel              The Braidwood Station, The Braidwood Station, Unit 2 dimensions must be        Unit 1 SG vessel          SG vessel dimensions are as within 10% percent of      dimensions are as        follows:
the upper and lower        follows:
bounds of the values
* SG Lower Head diameter provided in the table in
* SG  Lower Head            = 135.38 inches Section 9.4.3 of                diameter = 136.88
* SG Upper Shell diameter Reference [A1].                inches                    = 176.25 inches
* SG Upper Shell diameter = 176.75 These dimensions are within inches                10% of those specified in Table 9-2 in Section 9.4.3 of These dimensions are      Reference [A1] for within 10% of those        Westinghouse plants.
specified in Table 9-2 in Section 9.4.3 of Reference [A1] for Westinghouse plants.
The component must        As shown in Table A3, the Braidwood Station, Units 1 experience transients      and 2 transients and number of cycles projected to and cycles bounded        occur over a 60-year life are bounded by those shown in by those shown in          Table 5-7 of Reference [A1] for Westinghouse plants.
Table 5- 7 of Reference [A1] over a 60-year operating life.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 32 of 98)
Table A2. Applicability of Reference [A1] Representative Analyses to Braidwood Station, Units 1 and 2 Items Numbers C1.10, C1.20 and C1.30 (SG Secondary Side Shell Welds)
Requirement from                Applicability to            Applicability to Category Reference [A1]              Braidwood, Unit 1          Braidwood, Unit 2 General      The Loss of Power              Braidwood Station, Units 1 and 2 has not Requirements    transient (involving            experienced a loss of power transient resulting in unheated auxiliary              unheated auxiliary feedwater being introduced into feedwater being                a hot SG that has been boiled dry following introduced into a hot SG        blackout, resulting in thermal shock of portion of that has been boiled dry        the vessel.
following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 33 of 98)
Requirement from              Applicability to        Applicability to Category Reference [A1]            Braidwood, Unit 1      Braidwood, Unit 2 The materials of the SG      The Braidwood          The Braidwood Station, shell, FW nozzles, and        Station, Unit 1 SG    Unit 2, SG tube plate is MS nozzles must be            tubesheet, shell cone, fabricated from SA-508, low alloy ferritic steels    and steam drum head    Class 2a material and which conform to the          are fabricated of SA-  the SG barrels, requirements of ASME          508, Class 3 material, transition cone, and Code, Section XI,            and the SG steam      upper head are Appendix G, Paragraph        drum shell and        fabricated from SA-533, G-2110.                      secondary shell are    Grade A, Class 2.
fabricated from SA-    Material SA-508, Class 533, Grade B, Class 1  2a conforms to the material. These        requirements of ASME materials conform to  Code, Section XI, the requirements of    Appendix G, Paragraph ASME Code, Section    G- 2110.
XI, Appendix G, Paragraph G- 2110. The material properties in SA-533, Grade A, Class 2 material are identical compared to the SA-533, Grade B Class 1 material used in the FEA in the EPRI report as is shown Table 5-2 of Reference
[A1]. Appendix G, Table G-2110-1, of the 2019 Edition of ASME Section XI acknowledges that Figure G-2210-1 is applicable to material SA-533, Grade A, Class 2. Therefore, it can be concluded that the EPRI report is applicable to material SA-533, Grade A, Class 2.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 34 of 98)
Requirement from              Applicability to            Applicability to Category Reference [A1]            Braidwood, Unit 1          Braidwood, Unit 2 Specific      The weld                      The Braidwood              The Braidwood Station, Requirements    configurations must          Station, Unit 1 weld      Unit 2 weld conform to those              configurations are        configurations are shown in Figure 1-7          shown in Figure A2        shown in Figure A3 and and Figure 1-8 of            and conform to Figures    conforms to Figures 1-Reference [A1].              1-7 and 1-8 of            7 and 1-8 of Reference Reference [A1].            [A1].
The SG vessel                The Braidwood Station,    The Braidwood Station, dimensions must be            Unit 1, SG vessel          Unit 2 SG vessel within 10% percent of        dimensions are as          dimensions are as the upper and lower          follows:                  follows:
bounds of the values provided in the table
* SG Lower Head
* SG Lower Head in Section 9.4.4 of                diameter = 136.88          diameter Reference [A1].                    inches                    = 135.38 inches
* SG Upper Shell
* SG Upper Shell diameter = 176.75          diameter inches                    = 176.25 inches These dimensions are      These dimensions are within 10% of those        within 10% of those specified in Table 9-3 in specified in Table 9-3 in Section 9.4.4 of          Section 9.4.4 of Reference [A1] for        Reference [A1] for Westinghouse plants.      Westinghouse plants.
The component must            As shown in Table A4, the Braidwood Station, Units experience transients        1 and 2, transients and number of cycles projected and cycles bounded            to occur over a 60-year life are bounded by those by those shown in            shown in Table 5-9 of Reference [A1].
Table 5-9 of Reference [A1] over a 60-year operating life.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 35 of 98)
 
==REFERENCES:==
 
A1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906, ADAMS Accession No. ML20225A141.
A2. Braidwood, Unit 1, Drawing 7720E038, Section XI Preservice U/T.
A3. Braidwood, Unit 2, Drawing 1103J99, Sheet 1, Steam Generator General Arrangement D5-3.
A4. Braidwood, Unit 1, Drawing 1SG-05, Inservice Inspection Drawing for Replacement Steam Generator 1RC01BA - Loop 1, Unit 1.
A5. Braidwood, Unit 2, Drawing 2SG-01, Spec L2907, Inspection Identification Drawing for Inservice Inspection of Steam Generator 2RC01BA - Loop #1, Unit 2.
A6. Braidwood, Unit 1, Design Specification 18-1229648, Certified Design Specification for Replacement Steam Generator, Byron and Braidwood Stations Unit 1.
A7. Braidwood, Unit 2, Design Specification G-953431, Reactor Coolant System - Model D-5 Steam Generator.
A8. AREVA Specification 18-1229648-009 (MUR), Certified Design Specification for Replacement Steam Generator Byron and Braidwood Stations Unit 1.
A9. Framatone 32-500-7271-00, Byron/Braidwood RSG Design Transients - Power Uprate.
A10. WCAP-11388, Volume 2, Byron Units 1 and 2 and Braidwood Units 1 and 2, T-Hot Reduction Program Engineering Report.
A11. Design Specification 953442, Rev 8, Model D5 Steam Generator (SG) Stress Report.
A12. CN-OPES(99)-061, Revision 0, Power Uprate - Design Transient Revisions for BY/BR Uprating.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 36 of 98)
Figure A1. Braidwood Station, Units 1 and 2 SG Primary Side Head to Tubesheet Weld Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 37 of 98)
Figure A2. Braidwood Station, Unit 1, SG Secondary Side Shell Weld Configurations
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 38 of 98)
Figure A3. Braidwood Station, Unit 2 SG Secondary Side Shell Weld Configurations
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 39 of 98)
Figure A4. Braidwood, Unit 1, Steam Generator General Outline
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 40 of 98)
Figure A5. Braidwood, Unit 2, Steam Generator General Outline
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 41 of 98)
Table A3. Transient Cycles for PWR SG Primary-Side Head Welds for Braidwood Station, Units 1 and 2 in Comparison to Table 5-7 of Reference [A1]
Maximum        Minimum        Maximum        Minimum        Maximum          Minimum 60-Year Transient              THOT          THOT          TCOLD            TCOLD        Pressure        Pressure Cycles
(°F)          (°F)            (°F)            (°F)        (psig)          (psig)
Heatup/Cooldown From Reference            545            70            545              70            2235              0                300
[A1]
Heatup/Cooldown for Braidwood                                                                                                          72/70 [Unit 1]
5573            120            5573            120          22503            467 Station, Units 1 and                                                                                                      79/94 [Unit 2]
2 [A10, A12]
Plant Loading/
Unloading from            610            550            550              545            2300            2300              5000 Reference [A1]
Plant Loading/
Unloading for 2661/2182 [Unit 1]
Braidwood Station,        618.43          566.5          558.13          538.23          23703          22953 2201/1862 [Unit 2]
Units 1 and 2 [A10, A12]
Reactor Trip from 615            530            565              530            2435            1700                360 Reference [A1]
Reactor Trip for Braidwood Station,                                                                                                        11 [Unit 1]
618.43          4353          555.2            538.2          2250          15823 Units 1 and 2 [A10,                                                                                                        69 [Unit 2]
A12]
Note:
: 1.      60-Year Cycle counts for Plant Loading conservatively includes all events for transients Plant Loading at 5%/min, Plant Loading, 0% - 15%, Cold Turb/Gen, and Plant Loading, 0% - 15%, Hot Turb/Gen.
: 2.      60-Year Cycle counts for Plant Unloading conservatively includes all events for transients Plant Unloading at 5%/min and Plant Unloading, 15% - 0%.
: 3.      See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 42 of 98)
Table A4. Transient Cycles for PWR SG Secondary-Side Vessel Welds for Braidwood Station, Units 1 and 2 in Comparison to Table 5-9 of Reference [A1]
Maximum          Minimum Maximum Tss      Minimum Tss Transient                                                        Pressure          Pressure        60-Year Cycles
(°F)              (°F)
(psig)              (psig)
Heatup/Cooldown from 545                70                1000                0                  300 Reference [A1]
Heatup/Cooldown for 72/70 [Unit 1]
Braidwood Station, Units              5573              120              10773                  0 79/94 [Unit 2]
1 and 2 [A8, A10]
Plant Loading/ Unloading 545                540                1000                1000                5000 from Reference [A1]
Plant Loading/
Unloading for Braidwood          555.5 [Unit 1]3    528.8 [Unit 1]3    1077 [Unit 1]3,4  1077 [Unit 1]4  2661/2182 [Unit 1]
Station, Units 1 and 2          554.3 [Unit 2]3  524.4 [Unit 2]3    1077 [Unit 2]3,4  1077 [Unit 2]4  2201/1862 [Unit 2]
[A8, A9]
Reactor Trip from 555                530                1130                1000                360 Reference [A1]
Reactor Trip for 557.9 [Unit 1]3  460.9 [Unit 1]3      1092 [Unit 1]    995 [Unit 1]3,6      11 [Unit 1]
Braidwood Station, Units 553.8 [Unit 2]    458.8 [Unit 2]3      1092 [Unit 2]5    860 [Unit 2]3,6      69 [Unit 2]
1 and 2 [A8, A9]
Note:
: 1.        60-Year Cycle counts for Plant Loading conservatively includes all events for transients Plant Loading at 5%/min and Plant Loading, 0% - 15%.
: 2.        60-Year Cycle counts for Plant Unloading conservatively includes all events for transients Plant Unloading at 5%/min and Plant Unloading, 15% - 0%.
: 3.        See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
: 4.        Maximum and Minimum Pressure values for the Plant Loading/Unloading Transient could not be located in design documentation for the Secondary Side of the Vessel. Since the Maximum and Minimum TSS for the Heatup/Cooldown Transient are similar to the Loading/Unloading Transient, the same pressure was assumed.
This is the basis for the 1077 psig values listed in the table.
: 5.        Maximum Pressure for the Reactor Trip transient for Unit 2 could not be located in design documentation. Due to the similar Maximum Tss values for the Reactor Trip Transient between Units 1 and 2, the same Maximum Pressure was assumed for Unit 2.
: 6.        Minimum Pressure for the Reactor Trip transient for Unit 1 and 2 could not be located in design documentation.
Plant operating data for Byron and Braidwood was reviewed, and the minimum pressure corresponding to 100% reactor power was used. This pressure corresponds to the minimum pressure when a reactor trip transient would initiate at 100% power. Due to similarities in SG design and operational strategies between the stations, the same values were used for Braidwood Unit 1 and Byron Unit 1 and also for Braidwood Unit 2 and Byron Unit 2.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 43 of 98)
Plant-Specific Applicability for Byron Section 9 of Reference [A13] provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Byron, Units 1 and 2, is provided in Tables A5 through A8.
Tables A5 through A8 indicate that all plant-specific requirements are met for Byron, Units 1 and
: 2. Therefore, the results and conclusions of the EPRI report are applicable to Byron, Units 1 and 2.
Table A5. Applicability of Reference [A13] Representative Analyses to Byron, Units 1 and 2 Items Number B2.40 (SG Primary Side Shell Welds)
Requirement from              Applicability to        Applicability to Byron, Category Reference [A13]                Byron, Unit 1                      Unit 2 General      The Loss of Power            Byron Station, Units 1 and 2 has not experienced a Requirements    transient (involving        loss of power transient resulting in unheated auxiliary unheated auxiliary          feedwater being introduced into a hot SG that has feedwater being              been boiled dry following blackout, resulting in thermal introduced into a hot        shock of portion of the vessel.
SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 44 of 98)
The materials of the      The Byron Station,    The Byron Station, Unit 2 SG shell, FW nozzles,    Unit 1, SG vessel      SG channel head is and MS nozzles must      primary head and      fabricated of SA-216, Grade be low alloy ferritic    tubesheet are          WCC, and the SG tube steels which conform      fabricated of SA-508,  plate is fabricated from SA-to the requirements of    Class 3 material. This 508, Class 2a material.
ASME Code, Section        material conforms to  Material SA-508, Class 2a XI, Appendix G,          the requirements of    conforms to the Paragraph G- 2110.        ASME Code, Section    requirements of ASME Code, Section XI, Appendix XI, Appendix G, G, Paragraph G- 2110.
Paragraph G-2110.
SA-216 Gr. WCC material is not specifically listed in ASME Code, Section XI, Appendix G. However, SA-216 Grade WCC material has similar toughness and chemical composition to SA-533, Grade B, Class 1, SA-508-1, SA-508-2, and SA-508-3, it has a specified minimum yield strength at room temperature of 40 ksi (which is less than 50 ksi),
and the RTNDT values for the Byron Unit 2 steam generators channel head material was required to be 60°F or less (so the RTNDT of 60°F used in the EPRI report is bounding). Also, thermal transient stresses in SA-216 Grade WCC material are lower compared to the SA-533 Grade B Class 1 material used in the FEA in the EPRI report. Therefore, by comparison, SA-216 Gr.
WCC material is consistent with the requirements of ASME Code, Section XI, Appendix G and satisfies the requirements for application of the EPRI report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 45 of 98)
Requirement from            Applicability to        Applicability to Byron, Category Reference [A13]            Byron, Unit 1                    Unit 2 Specific    The weld                  The Byron Station,        The Byron Station, Unit 2 Requirements    configurations must        Unit 1 weld                weld configuration is conform to those          configuration is shown shown in Figure A6 and shown in Figures 1-1      in Figure A6 and          conforms to Figure 1-2 of and Figure 1-2 of          conforms to Figure 1-      Reference [A13].
Reference [A13].          2 of Reference [A13].
The SG vessel              The Byron Station, Unit The Byron Station, Unit 2 SG dimensions must be        1 SG vessel dimensions vessel dimensions are as within 10% percent of      are as follows:          follows:
the upper and lower bounds of the values
* SG Lower Head
* SG Lower Head diameter provided in the table in      diameter = 136.88          = 135.58 inches Section 9.4.3 of              inches
* SG Upper Shell diameter Reference [A13].
* SG Upper Shell            = 176.5 inches diameter = 176.75 inches                These dimensions are within 10% of those specified in These dimensions are      Table 9-2 in Section 9.4.3 of within 10% of those      Reference [A13] for specified in Table 9-2    Westinghouse plants.
in Section 9.4.3 of Reference [A13] for Westinghouse plants.
The component must        As shown in Table A7, the Byron Station, Units 1 and 2 experience transients      transients and number of cycles projected to occur over and cycles bounded        a 60-year life are bounded by those shown in Table 5-7 by those shown in          of Reference [A13] for Westinghouse plants.
Table 5- 7 of Reference [A13] over a 60-year operating life.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 46 of 98)
Table A6. Applicability of Reference [A13] Representative Analyses to Byron, Units 1 and 2 Items Numbers C1.10, C1.20 and C1.30 (SG Secondary Side Shell Welds)
Requirement from                Applicability to            Applicability to Category Reference [A13]                Byron, Unit 1                Byron, Unit 2 General      The Loss of Power              Byron Station, Units 1 and 2 has not experienced a Requirements  transient (involving          loss of power transient resulting in unheated auxiliary unheated auxiliary            feedwater being introduced into a hot SG that has feedwater being                been boiled dry following blackout, resulting in introduced into a hot SG      thermal shock of portion of the vessel.
that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this report's guidance.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 47 of 98)
Requirement from              Applicability to        Applicability to Category Reference [A13]                Byron, Unit 1            Byron, Unit 2 The materials of the SG      The Byron Station, Unit  The Byron Station, Unit shell, FW nozzles, and      1 SG tubesheet, shell    2, SG tube plate is MS nozzles must be          cone, and steam drum    fabricated from SA-508, low alloy ferritic steels    head are fabricated of  Class 2a material and the which conform to the        SA-508, Class 3          SG barrels, transition requirements of ASME        material, and the SG    cone, and upper head Code, Section XI,            steam drum shell and    are fabricated from SA-Appendix G, Paragraph        secondary shell are      533, Grade A, Class 2.
G-2110.                      fabricated from SA-533,  Material SA-508, Class Grade B, Class 1        2a conforms to the material. These          requirements of ASME materials conform to the Code, Section XI, requirements of ASME    Appendix G, Paragraph Code, Section XI,        G- 2110.
Appendix G, Paragraph G- 2110.                The material properties in SA-533, Grade A, Class 2 material are identical compared to the SA-533, Grade B Class 1 material used in the FEA in the EPRI report as is shown Table 5-2 of Reference
[A13]. Appendix G, Table G-2110-1, of the 2019 Edition of ASME Section XI acknowledges that Figure G-2210-1 is applicable to material SA-533, Grade A, Class 2.
Therefore, it can be concluded that the EPRI report is applicable to material SA-533, Grade A, Class 2.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 48 of 98)
Requirement from            Applicability to            Applicability to Category Reference [A13]              Byron, Unit 1              Byron, Unit 2 Specific    The weld                  The Byron Station, Unit The Byron Station, Unit 2 Requirements    configurations must        1 weld configurations        weld configurations are conform to those          are shown in Figure A7 shown in Figure A8 and shown in Figure 1-7        and conform to Figures conforms to Figures 1-7 and Figure 1-8 of          1-7 and 1-8 of              and 1-8 of Reference Reference [A13].          Reference [A13].            [A13].
The SG vessel              The Byron Station, Unit 1 The Byron Station, Unit 2 dimensions must be          SG vessel dimensions        SG vessel dimensions are within 10% percent of      are as follows:            as follows:
the upper and lower bounds of the values
* SG Lower Head
* SG Lower Head provided in the table            diameter = 136.88          diameter in Section 9.4.4 of              inches                    = 135.58 inches Reference [A13].
* SG Upper Shell
* SG Upper Shell diameter = 176.75          diameter inches                    = 176.5 inches These dimensions are        These dimensions are within 10% of those          within 10% of those specified in Table 9-3 in    specified in Table 9-3 in Section 9.4.4 of            Section 9.4.4 of Reference Reference [A13] for          [A13] for Westinghouse Westinghouse plants.        plants.
The component must        As shown in Table A8, the Byron Station, Units 1 and experience transients      2 transients and number of cycles projected to occur and cycles bounded          over a 60-year life are bounded by those shown in by those shown in          Table 5-9 of Reference [A13].
Table 5-9 of Reference [A13] over a 60-year operating life.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 49 of 98)
 
==REFERENCES:==
 
A13. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906, ADAMS Accession No. ML20225A141.
A14. Byron, Unit 1, Drawing 7720E038, Section XI Preservice U/T.
A15. Byron, Unit 2, Drawing 1512E55, Sheet 1, Steam Generator Model D5 As Built Dimensions.
A16. Byron, Unit 1, Drawing 1SG-1-ISI, Inspection Identification Drawing for Inservice Inspection for Replacement Steam Generator No. 1RC01BA.
A17. Byron, Unit 2, Drawing 2SG-1-ISI, Inspection Identification Drawing for Inservice Inspection of Steam Generator No. 2RC01BA.
A18. Byron, Unit 1, Specification 18-1229648, Certified Design Specification for Replacement Steam Generator, Byron and Braidwood Stations Unit 1.
A19. Byron, Unit 2, Design Specification G-953431, Reactor Coolant System - Model D-5 Steam Generator.
A20. Byron Work Order 05140866-01, LR-Cycle Counting (Data through 3/31/2021)
A21. AREVA Specification 18-1229648-009 (MUR), Certified Design Specification for Replacement Steam Generator Byron and Braidwood Stations Unit 1.
A22. Framatone 32-500-7271-00, Byron/Braidwood RSG Design Transients - Power Uprate.
A23. WCAP-11388, Volume 2, Byron Units 1 and 2 and Braidwood Units 1 and 2, T-Hot Reduction Program Engineering Report.
A24. Design Specification 953442, Rev 8, Model D5 Steam Generator (SG) Stress Report.
A25. CN-OPES(99)-061, Revision 0, Power Uprate - Design Transient Revisions for BY/BR Uprating.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 50 of 98)
Figure A6. Byron Station, Units 1 and 2 SG Primary Side Head to Tubesheet Weld Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 51 of 98)
Figure A7. Byron, Unit 1, SG Secondary Side Shell Weld Configurations
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 52 of 98)
Figure A8. Byron, Unit 2 SG Secondary Side Shell Weld Configurations
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 53 of 98)
Figure A9. Byron, Unit 1, Steam Generator General Outline
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 54 of 98)
Figure A10. Byron, Unit 2, Steam Generator General Outline
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 55 of 98)
Table A7. Transient Cycles for PWR SG Primary-Side Head Welds for Byron, Units 1 and 2 in Comparison to Table 5-7 of Reference [A13]
Maximum        Minimum        Maximum          Minimum        Maximum          Minimum Transient              THOT          THOT          TCOLD          TCOLD        Pressure        Pressure        60-Year Cycles
(°F)          (°F)            (°F)          (°F)          (psig)          (psig)
Heatup/Cooldown From Reference            545            70            545              70            2235              0                  300
[A13]
Heatup/Cooldown for Byron Station,                                                                                                        66/66 [Unit 1]
5573          120            5573            120            22503            467 Units 1 and 2 [A20,                                                                                                      62/62 [Unit 2]
A23, A25]
Plant Loading/
Unloading from          610            550            550            545            2300            2300                5000 Reference [A13]
Plant Loading/
Unloading for 1811/1882 [Unit 1]
Byron Station,          618.43        566.5          558.13          538.23          23703          22953 1791/1792 [Unit 2]
Units 1 and 2 [A20, A23, A25]
Reactor Trip from 615            530            565            530            2435            1700                360 Reference [A13]
Reactor Trip for Byron Station,                                                                                                            9 [Unit 1]
618.43          4353          555.2          538.2            2250          15823 Units 1 and 2 [A20,                                                                                                        13 [Unit 2]
A23, A25]
Note:
: 1.      60-Year Cycle counts for Plant Loading conservatively includes all events for transients Plant Loading at 5%/min, Plant Loading, 0% - 15%, Cold Turb/Gen, and Plant Loading, 0% - 15%, Hot Turb/Gen.
: 2.      60-Year Cycle counts for Plant Unloading conservatively includes all events for transients Plant Unloading at 5%/min and Plant Unloading, 15% - 0%.
: 3.      See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 56 of 98)
Table A8. Transient Cycles for PWR SG Secondary-Side Vessel Welds for Byron, Units 1 and 2 in Comparison to Table 5-9 of Reference [A13]
Maximum Tss        Minimum Tss          Maximum              Minimum Transient                                                                                              60-Year Cycles
(°F)              (°F)          Pressure (psig)    Pressure (psig)
Heatup/Cooldown 545                70                1000                  0                    300 from Reference [A13]
Heatup/Cooldown for 66/66 [Unit 1]
Byron Station, Units 1            5573                120                10773                  0 62/62 [Unit 2]
and 2 [A20, A21, A23]
Plant Loading/
Unloading from                545                540                1000                1000                  5000 Reference [A13]
Plant Loading/
Unloading for Byron        555.5 [Unit 1]3    525.8 [Unit 1]3      1077 [Unit 1]3,4    1077 [Unit 1]4      1811/1882 [Unit 1]
Station, Units 1 and 2      554.3 [Unit 2]3    524.4 [Unit 2]3      1077 [Unit 2]3,4    1077 [Unit 2]4      1791/1792 [Unit 2]
[A20, A21, A22]
Reactor Trip from 555                530                1130                1000                    360 Reference [A13]
Reactor Trip for Byron 557.9 [Unit 1]3    460.9 [Unit 1]      1092 [Unit 1]      995 [Unit 1]3,6          9 [Unit 1]
Station, Units 1 and 2 553.8 [Unit 2]      458.8 [Unit 2]      1092 [Unit 2]5    860 [Unit 2]3,6          13 [Unit 2]
[A20, A21, A22]
Note:
: 1.      60-Year Cycle counts for Plant Loading conservatively includes all events for transients Plant Loading at 5%/min and Plant Loading, 0% - 15%.
: 2.      60-Year Cycle counts for Plant Unloading conservatively includes all events for transients Plant Unloading at 5%/min and Plant Unloading, 15% - 0%.
: 3.      See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
: 4.      Maximum and Minimum Pressure values for the Plant Loading/Unloading Transient could not be located in design documentation for the Secondary Side of the Vessel. Since the Maximum and Minimum TSS for the Heatup/Cooldown Transient are similar to the Loading/Unloading Transient, the same pressure was assumed.
This is the basis for the listed 1077 psig values listed in the table.
: 5.      Maximum Pressure for the Reactor Trip transient for Unit 2 could not be located in design documentation. Due to the similar Maximum Tss values for the Reactor Trip Transient between Units 1 and 2, the same Maximum Pressure was assumed for Unit 2.
: 6.      Minimum Pressure for the Reactor Trip transient for Unit 1 and 2 could not be located in design documentation.
Plant operating data for Byron and Braidwood was reviewed, and the minimum pressure corresponding to 100% reactor power was used. This pressure corresponds to the minimum pressure when a reactor trip transient would initiate at 100% power. Due to similarities in SG design and operational strategies between the stations, the same values were used for Braidwood Unit 1 and Byron Unit 1 and also for Braidwood Unit 2 and Byron Unit 2.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 57 of 98)
Plant-Specific Applicability for Calvert Cliffs Section 9 of Reference [A26] provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Calvert Cliffs, Units 1 and 2, is provided in Tables A9 through A13.
Tables A9 through A13 indicate that all plant-specific requirements are met for Calvert Cliffs, Units 1 and 2. Therefore, the results and conclusions of the EPRI report are applicable to Calvert Cliffs, Units 1 and 2.
Table A9. Applicability of Reference [A26] Representative Analyses to Calvert Cliffs, Units 1 and 2 Items Number B2.40 (SG Primary Side Shell Welds)
Applicability to Calvert Category          Requirement from Reference [A26]
Cliffs, Units 1 and 2 General          The Loss of Power transient (involving          Calvert Cliffs, Units 1 and 2 Requirements        unheated auxiliary feedwater being              has not experienced a loss introduced into a hot SG that has been          of power transient resulting boiled dry following blackout, resulting in      in unheated auxiliary thermal shock of portion of the vessel) is not  feedwater being introduced considered in this evaluation due to its rarity. into a hot SG that has been In the event that such a significant thermal    boiled dry following event occurs at a plant, its impact on the KIC  blackout, resulting in (material fracture toughness) value may          thermal shock of portion of require more frequent examinations and          the vessel.
other plant actions outside the scope of this report's guidance.
The materials of the SG shell, FW nozzles,      The Calvert Cliffs, Units 1 and MS nozzles must be low alloy ferritic        and 2 SG vessel primary steels which conform to the requirements of      head and tubesheet are ASME Code, Section XI, Appendix G,              fabricated of SA-508, Class Paragraph G- 2110.                              3a material. This material conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G- 2110.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 58 of 98)
Applicability to Calvert Category        Requirement from Reference [A26]
Cliffs, Units 1 and 2 Specific      The weld configurations must conform to        The Calvert Cliffs, Units 1 Requirements    those shown in Figures 1-1 and Figure 1-2 of  and 2 weld configurations Reference [A26].                              are shown in Figure A11 and conform to Figure 1-2 of Reference [A26].
The SG vessel dimensions must be within        The Calvert Cliffs, Units 1 and 10% percent of the upper and lower bounds      2 SG vessel dimensions are of the values provided in the table in Section as follows:
9.4.3 of Reference [A26].
* SG Lower Head diameter
                                                                      = 169 inches
* SG Upper Shell diameter
                                                                      = 239.63 inches These dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference
[A26] for CE plants.
The component must experience transients      As shown in Table A12, the and cycles bounded by those shown in          Calvert Cliffs, Units 1 and 2 Table 5- 7 of Reference [A26] over a 60-year  transients and number of operating life.                                cycles projected to occur over a 60-year life are bounded by those shown in Table 5-7 of Reference [A26] for CE plants.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 59 of 98)
Table A10. Applicability of Reference [A26] Representative Analyses to Calvert Cliffs, Units 1 and 2 Item Number B3.130 (SG Primary Inlet/Outlet Nozzles)
Applicability to Calvert Category          Requirement from Reference [A26]
Cliffs, Units 1 and 2 General        The Loss of Power transient (involving          Calvert Cliffs, Units 1 and 2 Requirements    unheated auxiliary feedwater being              has not experienced a loss introduced into a hot SG that has been boiled    of power transient resulting dry following blackout, resulting in thermal    in unheated auxiliary shock of portion of the vessel) is not          feedwater being introduced considered in this evaluation due to its rarity. into a hot SG that has been In the event that such a significant thermal    boiled dry following event occurs at a plant, its impact on the KIC  blackout, resulting in (material fracture toughness) value may          thermal shock of portion of require more frequent examinations and          the vessel.
other plant actions outside the scope of this report's guidance.
The materials of the SG shell and nozzles        The Calvert Cliffs, Units 1 must be low alloy ferritic steels which          and 2 SG vessel primary conform to the requirements of ASME Code,        head and primary inlet and Section XI, Appendix G, Paragraph G-2110.        outlet nozzles are fabricated of SA- 508, Class 3a material. This material conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110.
Specific      The weld configurations must conform to          The Calvert Cliffs, Units 1 Requirements      those shown in Figures 1-3 through 1-5 of        and 2 weld configurations Reference [A26].                                are shown in Figure A12 and conform to Figure 1-5 of Reference [A26].
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 60 of 98)
Applicability to Calvert Category        Requirement from Reference [A26]
Cliffs, Units 1 and 2 The piping attached to the primary inlet and  The Calvert Cliffs, Units 1 outlet nozzles (RCS piping) for the various  and 2 primary inlet/outlet designs must be within 10% percent of the    nozzle and piping values provided in the table in Section 9.4.2 dimensions are as follows:
of Reference [A26].
* SG primary side inlet nozzle= 42 inches NPS
* SG primary side outlet nozzle= 30 inches NPS These dimensions are within 10% of those specified in Table 9-1 in Section 9.4.2 of Reference [A26] for CE plants.
The component must experience transients      As shown in Table A12, the and cycles bounded by those shown in Table    Calvert Cliffs, Units 1 and 2 5-7 of Reference [A26] over a 60-year        transients and number of operating life.                              cycles projected to occur over a 60-year life are bounded by those shown in Table 5-7 of Reference
[A26] for CE plants.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 61 of 98)
Table A11. Applicability of Reference [A26] Representative Analyses to Calvert Cliffs, Units 1 and 2 Items Numbers C1.10, C1.20 and C1.30 (SG Secondary Side Shell Welds)
Applicability to Calvert Category        Requirement from Reference [A26]
Cliffs, Units 1 and 2 General      The Loss of Power transient (involving          Calvert Cliffs, Units 1 and 2 Requirements    unheated auxiliary feedwater being              has not experienced a loss introduced into a hot SG that has been          of power transient resulting boiled dry following blackout, resulting in      in unheated auxiliary thermal shock of portion of the vessel) is not  feedwater being introduced considered in this evaluation due to its rarity. into a hot SG that has been In the event that such a significant thermal    boiled dry following blackout, event occurs at a plant, its impact on the KIC  resulting in thermal shock of (material fracture toughness) value may          portion of the vessel.
require more frequent examinations and other plant actions outside the scope of this report's guidance.
The materials of the SG shell, FW nozzles,      The Calvert Cliffs, Units 1 and MS nozzles must be low alloy ferritic        and 2, SG vessel original steels which conform to the requirements        upper head, cone segment of ASME Code, Section XI, Appendix G,            and steam drum shell are Paragraph G-2110.                                fabricated from SA-533, Grade B, Class 1 material, and the SG replaced shell cone, secondary shell assembly, and tubesheet are fabricated from SA-508, Class 3a. These materials conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 62 of 98)
Applicability to Calvert Category        Requirement from Reference [A26]
Cliffs, Units 1 and 2 Specific    The weld configurations must conform to    The Calvert Cliffs, Units 1 and Requirements  those shown in Figure 1-7 and Figure 1-8  2 weld configurations are of Reference [A26].                        shown in Figure A13, and conform to Figures 1-7 and 1-8 of Reference [A26].
The SG vessel dimensions must be within    The Calvert Cliffs, Units 1 and 10% percent of the upper and lower        2 SG vessel dimensions are bounds of the values provided in the table as follows:
in Section 9.4.4 of Reference [A26].
* SG Lower Head diameter
                                                                  = 169 inches
* SG Upper Shell diameter
                                                              *  = 239.63 inches These dimensions are within10% of those specified in Table 9-3 in Section 9.4.4 of Reference [A26] for CE plants.
The component must experience              As shown in Table A13, the transients and cycles bounded by those    Calvert Cliffs, Units 1 and 2 shown in Table 5-9 of Reference [A26]      transients and number of over a 60-year operating life.            cycles projected to occur over a 60-year life are bounded by those shown in Table 5-9 of Reference [A26].
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 63 of 98)
 
==REFERENCES:==
 
A26. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906, ADAMS Accession No. ML20225A141.
A27. Calvert Cliffs, Units 1 and 2, Drawing 12010-0007, Upper Shell Details and Assembly Steam Generator.
A28. Calvert Cliffs, Units 1 and 2, Drawing 12010A-0001SH0001, General Arrangement.
A29. Calvert Cliffs, Units 1 and 2, Drawing 12010A-0015SH0001, Section XI Pre-Service NDE Examination.
A30. Cliffs, Units 1 and 2, Figure A-4, S.G. Tube Sheet & Lower Head.
A31. Cliffs, Units 1 and 2, Figure B-1, Steam Generator Outline.
A32. Calvert Cliffs, Units 1 and 2, Specification 8067-31-2, Engineering Specification for a Steam Generator Assembly for Baltimore Gas & Electric Company Calvert Cliffs Station.
A33. ECP-18-000545, Fatigue Plant Transient Data Review Report for the Year of 2018.
A34. Design Specification SP-0811, Replacement Steam Generator Design Specification.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 64 of 98)
Figure A11. Calvert Cliffs, Units 1 and 2, SG Primary Side Head to Tubesheet Weld Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 65 of 98)
Figure A12. Calvert Cliffs, Units 1 and 2, SG Primary Side Inlet/Outlet Nozzle Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 66 of 98)
Figure A13. Calvert Cliffs, Units 1 and 2, SG Secondary Side Shell Welds Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 67 of 98)
Figure A14. Calvert Cliffs, Units 1 and 2, Steam Generator General Outline
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 68 of 98)
Table A12. Transient Cycles for PWR SG Primary-Side Head Welds for Calvert Cliffs, Units 1 and 2 in Comparison to Table 5-7 of Reference [A26]
Maximum        Minimum        Maximum        Minimum        Maximum          Minimum Transient              THOT          THOT          TCOLD          TCOLD        Pressure        Pressure      60-Year Cycles
(°F)          (°F)          (°F)            (°F)          (psig)          (psig)
Heatup/Cooldown From Reference            545            70            545              70            2235              0                300
[A26]
Heatup/Cooldown for Calvert Cliffs                                                                                                    149/139 [Unit 1]
535              70          535              70            2235            100 Station, Units 1 and                                                                                                  137/129 [Unit 2]
2 [A34]
Plant Loading/
Unloading from            610            550            550              545            2300            2300              5000 Reference [A26]
Plant Loading/
Unloading for 146/137 [Unit 1]
Calvert Cliffs          605            5481          550            5331            2298            22351 123/114 [Unit 2]
Station, Units 1 and 2 [A34]
Reactor Trip from 615            530            565              530            2435            1700              360 Reference [A26]
Reactor Trip for Calvert Cliffs                                                                                                        171 [Unit 1]
604            537            557              534            2240            1785 Station, Units 1 and                                                                                                    147 [Unit 2]
2 [A34]
Note:
: 1.      See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 69 of 98)
Table A13. Transient Cycles for PWR SG Secondary-Side Vessel Welds for Calvert Cliffs, Units 1 and 2 in Comparison to Table 5-9 of Reference [A26]
Maximum          Minimum        Maximum          Minimum Transient              Tss            Tss        Pressure          Pressure        60-Year Cycles
(°F)            (°F)          (psig)            (psig)
Heatup/Cooldown 545              70            1000                0                300 from Reference [A26]
Heatup/Cooldown for 149/139 [Unit 1]
Calvert Cliffs Station,      531              70              870                58 137/129 [Unit 2]
Units 1 and 2 [A32]
Plant Loading/
Unloading from            545            540            1000              1000              5000 Reference [A26]
Plant Loading/
Unloading for Calvert                                                                        146/137 [Unit 1]
532            526              890              845 Cliffs Station, Units 1                                                                      123/114 [Unit 2]
and 2 [A32]
Reactor Trip from 555            530            1130              1000                360 Reference [A26]
Reactor Trip for 1711 [Unit 1]
Calvert Cliffs Station,      540            532              942              8851 147 [Unit 2]
Units 1 and 2 [A32]
Note:
: 1.        See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 70 of 98)
Plant-Specific Applicability for Ginna Section 9 of Reference [A35] provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Ginna, is provided in Tables A14 through A17.
Tables A14 through A17 indicate that all plant-specific requirements are met for Ginna.
Therefore, the results and conclusions of the EPRI report are applicable to Ginna.
Table A14. Applicability of Reference [A35] Representative Analyses to Ginna Items Number B2.40 (SG Primary Side Shell Welds)
Category        Requirement from Reference [A35]                  Applicability to Ginna General        The Loss of Power transient (involving          Ginna has not experienced Requirements      unheated auxiliary feedwater being              a loss of power transient introduced into a hot SG that has been          resulting in unheated boiled dry following blackout, resulting in      auxiliary feedwater being thermal shock of portion of the vessel) is not  introduced into a hot SG considered in this evaluation due to its rarity. that has been boiled dry In the event that such a significant thermal    following blackout, resulting event occurs at a plant, its impact on the KIC  in thermal shock of portion (material fracture toughness) value may          of the vessel.
require more frequent examinations and other plant actions outside the scope of this report's guidance.
The materials of the SG shell, FW nozzles,      The Ginna SG vessel and MS nozzles must be low alloy ferritic        primary head and steels which conform to the requirements of      tubesheet are fabricated of ASME Code, Section XI, Appendix G,              SA-508, Class 3 material.
Paragraph G- 2110.                              This material conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G- 2110.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 71 of 98)
Category        Requirement from Reference [A35]                Applicability to Ginna Specific      The weld configurations must conform to        The Ginna weld Requirements    those shown in Figures 1-1 and Figure 1-2 of  configuration is shown in Reference [A35].                              Figure A15 and conforms to Figure 1-2 of Reference
[A35].
The SG vessel dimensions must be within        The Ginna SG vessel 10% percent of the upper and lower bounds      dimensions are as follows:
of the values provided in the table in Section 9.4.3 of Reference [A35].
* SG Lower Head diameter
                                                                      = 130.63 inches
* SG Upper Shell diameter
                                                                      =166.38 inches These dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference
[A35] for Westinghouse plants.
The component must experience transients      As shown in Table A16, the and cycles bounded by those shown in          Ginna transients and number Table 5- 7 of Reference [A35] over a 60-year  of cycles projected to occur operating life.                                over a 60-year life are bounded by those shown in Table 5-7 of Reference [A35]
for Westinghouse plants.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 72 of 98)
Table A15. Applicability of Reference [A35] Representative Analyses to Ginna Items Numbers C1.10, C1.20 and C1.30 (SG Secondary Side Shell Welds)
Category        Requirement from Reference [A35]                Applicability to Ginna General    The Loss of Power transient (involving          Ginna has not experienced a Requirements  unheated auxiliary feedwater being              loss of power transient introduced into a hot SG that has been          resulting in unheated boiled dry following blackout, resulting in      auxiliary feedwater being thermal shock of portion of the vessel) is not  introduced into a hot SG that considered in this evaluation due to its rarity. has been boiled dry In the event that such a significant thermal    following blackout, resulting event occurs at a plant, its impact on the KIC  in thermal shock of portion (material fracture toughness) value may          of the vessel.
require more frequent examinations and other plant actions outside the scope of this report's guidance.
The materials of the SG shell, FW nozzles,      The Ginna SG vessel and MS nozzles must be low alloy ferritic        tubesheet, shell cone, and steels which conform to the requirements        steam drum head are of ASME Code, Section XI, Appendix G,            fabricated of SA-508, Class Paragraph G-2110.                                3 material, and the secondary shell and steam drum shell are fabricated of SA-533, Type B, Class 1 material. These materials conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 73 of 98)
Category          Requirement from Reference [A35]          Applicability to Ginna Specific      The weld configurations must conform to    The Ginna weld Requirements    those shown in Figure 1-7 and Figure 1-8  configurations are shown in of Reference [A35].                        Figure A16, and conform to Figures 1-7 and 1-8 of Reference [A35].
The SG vessel dimensions must be within    The Ginna SG vessel 10% percent of the upper and lower        dimensions are as follows:
bounds of the values provided in the table in Section 9.4.4 of Reference [A35].
* SG Lower Head diameter
                                                                  = 130.63 inches
* SG Upper Shell diameter
                                                                  =166.38 inches These dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [A35] for Westinghouse plants.
The component must experience              As shown in Table A17, the transients and cycles bounded by those    Ginna transients and number shown in Table 5-9 of Reference [A35]      of cycles projected to occur over a 60-year operating life.            over a 60-year life are bounded by those shown in Table 5-9 of Reference [A35].
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 74 of 98)
 
==REFERENCES:==
 
A35. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906, ADAMS Accession No. ML20225A141.
A36. Ginna, Drawing 7705E001, Steam Generator Arrangement - SGA.
A37. Ginna, Drawing 7705E158, Layout of Vessel Reference Points for Welds.
A38. Ginna, Drawing 7705E205, Pressure Boundary Welds Prep & Weld Nos.
A39. Design Specification 18-1224785-05, Design Specification for Replacement Steam Generator.
A40. Ginna, Fatigue Monitoring Program LR-FATM-PROGPLAN, Revision 6, License Renewal Aging Management Program Basis Document, December 20, 2017.
A41. Ginna, Calculation BWC-143O-B1, Revision 0, R.E. Ginna - Power Uprate versus Non Power Uprate Transient Comparison, December 10, 2004.
A42. Westinghouse Document Number CN-SCS-04-060, Revision 0, RGE Secondary Side Design Transients for 19.5% Uprated Conditions.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 75 of 98)
Figure A15. Ginna Primary Side Head to Tubesheet Weld Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 76 of 98)
Figure A16. Ginna SG Secondary Side Shell Welds Configuration
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 77 of 98)
Figure A17. Ginna Steam Generator General Outline
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 78 of 98)
Table A16. Transient Cycles for PWR SG Primary-Side Head Welds for Ginna in Comparison to Table 5-7 of Reference [A35]
Maximum        Minimum        Maximum          Minimum        Maximum          Minimum 60-Year Transient            THOT          THOT          TCOLD            TCOLD        Pressure        Pressure Cycles
(°F)          (°F)            (°F)            (°F)          (psig)          (psig)
Heatup/Cooldown          545            70            545              70            2235              0              300 From Reference
[A35]
Heatup/Cooldown          5472          100            5472            100            2235              0              153 for Ginna [A39, A40]
Plant Loading/          610            550            550              545            2300            2300          5000 Unloading from Reference [A35]
Plant Loading/
Unloading for                                Not typical operation, not counted - See Note 1 Ginna Reactor Trip from        615            530            565              530            2435            1700            360 Reference [A35]
Reactor Trip for        611.8          547          585.92            539.9          26382          1700            138 Ginna [A40, A41]
Note:
: 1.      Ginna is operated as a base load plant. Typically, it is continuously operated at 100% (nominal) steady-state power. The only exceptions to this are refueling outages, certain equipment tests or maintenance requiring down power, and unplanned transients.
: 2.      See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 79 of 98)
Table A17. Transient Cycles for PWR SG Secondary-Side Vessel Welds for Ginna in Comparison to Table 5-9 of Reference [A35]
Maximum          Minimum          Maximum        Minimum 60-Year Transient              Tss              Tss          Pressure      Pressure Cycles
(°F)            (°F)            (psig)        (psig)
Heatup/Cooldown            545              70              1000            0              300 from Reference [A35]
Heatup/Cooldown for          5472            100            1004.82            0              153 Ginna [A40, A42]
Plant Loading/          545              540              1000          1000            5000 Unloading from Reference [A35]
Plant Loading/
Not typical Operation, not counted - See Note 1 Unloading for Ginna Reactor Trip from          555              530              1130          1000              360 Reference [A35]
Reactor Trip for          5652            515.42            11792          7812              138 Ginna [A40, A41]
Note:
: 1.        Ginna is operated as a base load plant. Typically, it is continuously operated at 100% (nominal) steady-state power. The only exceptions to this are refueling outages, certain equipment tests or maintenance requiring down power, and unplanned transients.
: 2.        See the Stress Analysis discussion in Section 5 for a basis to the applicability of the EPRI report for transients having a maximum pressure/temperature higher than what is listed in the EPRI Report or a minimum pressure/temperature lower than what is listed in the EPRI Report.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 80 of 98)
APPENDIX B PLANT SPECIFIC INSPECTION HISTORY
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 81 of 98)
BRAIDWOOD, UNITS 1 AND 2, INSPECTION HISTORY Currently, the SG components for Braidwood, Units 1 and 2, satisfy all of the inspection requirements of ASME Code, Section XI, 2013 Edition. A summary of the inspection history for Item Numbers B2.40, C1.10, C1.20, and C1.30 for Braidwood, Units 1 and 2, is provided in Tables B1 through B4.
Table B1. Braidwood SG Primary Side Tubesheet-to-Shell Welds (B2.40)
Item                                                                            Exam Unit        Date          Interval/Period      Component ID                  Coverage No.                                                                          Results B2.40                9/2001 1                    2nd Interval/ 2nd Period 1SG-05-SGC-01        NRI        100%
(A1R09) 10/2010 1                    3rd Interval/ 1st Period 1SG-05-SGC-01        NRI        100%
(A1R15) 10/1997 2                    1st Interval/ 3rd Period 2SG-04-SGC-01        NRI1        99.2%
(A2R06) 4/1990 2                    1st Interval/ 3rd Period 2SG-01-SGC-01        NRI1        100%
(A2R06) 10/1994 2                    1st Interval/ 2nd Period 2SG-02-SGC-01        NRI1        100%
(A1R04) 11/2003 2                    2nd Interval/ 2nd Period 2SG-02-SGC-01        NRI1        100%
(A2R10) 10/2012 2                    3rd Interval/ 2nd Period 2SG-02-SGC-01        NRI1        99.7%
(A2R16) 10/1997 2                    1st Interval/ 3rd Period 2SG-03-SGC-01        NRI1        99.2%
(A2R06)
Note:
: 1.      ID Geometry (45 Degree) noted for welds 2SG-04-SGC-01, 2SG-01-SGC-01, 2SG-02-SGC-01, and 2SG-03-SGC-01.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 82 of 98)
Table B2. Braidwood SG Secondary Side Shell Welds (C1.10)
Item                                                                                        Exam Unit        Date          Interval/Period          Component ID                        Coverage No.                                                                                      Results C1.10                  4/2006 1                    2nd Interval/ 3rd Period    1SG-05-SGC-06              NRI          100%
(A1R12) 4/2018 1                      3rd Interval/ 3rd Period    1SG-05-SGC-06              NRI          100%
(A1R20) 4/1996 2                      1st Interval/ 3rd Period    2SG-03-SGC-03              NRI1          100%
(A2R05) 5/2008 2                    2nd Interval/ 3rd Period    2SG-03-SGC-03              NRI          99.9%
(A2R13) 5/2017 2                      3rd Interval/ 3rd Period    2SG-03-SGC-03              NRI1          100%
(A2R19) 3/1993 2                    1st Interval/ 2nd Period    2SG-03-SGC-05              NRI          100%
(A2R03) 4/2002 2                    2nd Interval/ 2nd Period    2SG-03-SGC-05              NRI          100%
(A2R09) 4/2011 2                      3rd Interval/ 1st Period    2SG-03-SGC-05              NRI          100%
(A2R15) 3/1993 2                    1st Interval/ 2nd Period    2SG-03-SGC-06              NRI2          100%
(A2R03) 4/2002 2                    2nd Interval/ 2nd Period    2SG-03-SGC-06              NRI2          100%
(A2R09) 4/2011 2                      3rd Interval/ 1st Period    2SG-03-SGC-06              NRI2          100%
(A2R15) 10/2021 2                      4th Interval/ 1st Period    2SG-03-SGC-06              NRI2          100%
(A2R22)
Note:
: 1.      Suspected slag noted (45 Degree) for 2SG-03-SGC-03 (A2R05) and 2SG-03-SGC-03 (A2R19).
: 2.      ID Geometry noted (45 Degree) for weld 2SG-03-SGC-06.
: 3.      Examination of the upper secondary shell - shell cone welds (1SG-05-SGC-05, 1SG-06-SGC-05, 1SG          SGC-05, and 1SG-08-SGC-05) were not examined during the Second or Third Intervals, due to the basis that these welds were not located at gross structural discontinuities as defined in ASME Section III NB-3213.2. The welds were removed from the Examination Category C-A Item Number C1.10 examination population as permitted by Table IWC-2500-1 Examination Category C-A Note 2.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 83 of 98)
Table B3. Braidwood SG Secondary Side Head Welds (C1.20)
Item                                                                      Exam Unit        Date          Interval/Period        Component ID          Coverage No.                                                                      Results C1.20                9/2001 1                    2nd Interval/ 2nd Period  1SG-05-SGC-08  NRI    100%
(A1R09) 4/2015 1                    3rd Interval/ 2nd Period  1SG-05-SGC-08  NRI    100%
(A1R18) 4/1996 2                    1st Interval/ 3rd Period  2SG-03-SGC-08  NRI1    100%
(A2R05) 5/2008 2                    2nd Interval/ 3rd Period  2SG-03-SGC-08  NRI1    100%
(A2R13) 5/2014 2                    3rd Interval/ 2nd Period  2SG-03-SGC-08  NRI1    100%
(A2R17)
Note:
: 1.      ID Geometry noted (45 Degree) for weld 2SG-03-SGC-08.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 84 of 98)
Table B4. Braidwood SG Secondary Side Tubesheet-to-Shell Welds (C1.30)
Item                                                                      Exam Unit        Date          Interval/Period      Component ID          Coverage No.                                                                    Results C1.30                9/2001 1                    2nd Interval/ 2nd Period  1SG-05-SGC-02  NRI      99%
(A1R09) 10/2010 1                    3rd Interval/ 1st Period 1SG-05-SGC-02  NRI    100%
(A1R15) 4/1990 2                    1st Interval/ 1st Period 2SG-01-SGC-02  NRI1  88.4%
(A2R01) 5/2008 2                    2nd Interval/ 3rd Period 2SG-01-SGC-02  NRI    88.4%
(A2R13) 10/2018 2                    3rd Interval/ 3rd Period 2SG-01-SGC-02  NRI    88.4%
(A2R20)
Note:
: 1.      ID Geometry noted (45 Degree) for weld 2SG-01-SGC-02 (A2R01).
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 85 of 98)
BYRON, UNITS 1 AND 2, INSPECTION HISTORY Currently, the SG components for Byron, Units 1 and 2, satisfy all of the inspection requirements of ASME Code, Section XI, 2007 Edition with the 2008 Addenda. A summary of the inspection history for Item Numbers B2.40, C1.10, C1.20, and C1.30 for Byron, Units 1 and 2, is provided in Tables B5 through B8.
Table B5. Byron SG Primary Side Tubesheet-to-Shell Welds (B2.40)
Item                                                                Exam Unit      Date    Interval/Period      Component ID                  Coverage No.                                                                Results B2.40              4/1999    2nd Interval/ 1st 1                                      1RC-01-BB/SGW-01    NRI            100%
(B1R09)          Period 9/2006      3rd Interval/ 1st 1                                      1RC-01-BB/SGW-01    NRI            100%
(B1R14)          Period 3/2017      4th Interval/ 1st 1                                      1RC-01-BB/SGW-01    NRI            100%
(B1R21)          Period 9/1996      1st Interval/ 3rd 2                                      2RC-01-BC/SGC-01  NRI1            100%
(B2R06)          Period 9/1990      1st Interval/ 1st                              90% - limited due 2                                      2RC-01-BA/SGC-01  NRI1 (B2R02)          Period                                      to weld pads 10/1999    2nd Interval/ 1st                                99% - limited due 2                                      2RC-01-BA/SGC-01  NRI1 (B2R08)          Period                                      to weld pads 10/2008      3rd Interval/ 1st                              98% - limited due 2                                      2RC-01-BA/SGC-01  NRI1 (B2R14)          Period                                      to weld pads 4/2019      4th Interval/ 1st                              98.76% - limited 2                                      2RC-01-BA/SGC-01  NRI1 (B2R21)          Period                                    due to weld pads 10/1993    1st Interval/ 2nd                                90% - limited due 2                                      2RC-01-BB/SGC-01  NRI1 (B2R04)          Period                                      to weld pads 8/1996      1 Interval/ 3rd st 2                                      2RC-01-BD/SGC-01  NRI1            100%
(B2R06)          Period Note:
: 1.      Noted ID Geometry.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 86 of 98)
Table B6. Byron SG Secondary Side Shell Welds (C1.10)
Item                    Interval/
Unit    Date                    Component ID            Exam Results            Coverage No.                      Period C1.10                                                                                    99.5% -
3/2005  2nd Interval/                                                limited due to 1                            1RC-01-BB/SGW-06                NRI (B1R13)    3rd Period                                                      adjacent nozzles 3/2014  3rd Interval/
1                            1RC-01-BB/SGW-06                NRI                  100%
(B1R19)    3rd Period 3/2023  4th Interval/
1                            1RC-01-BB/SGW-06                NRI                  100%
(B1R25)    3rd Period 3/2023  4th Interval/
1                            1RC-01-BC/SGW-06                NRI                  100%
(B1R25)    3rd Period Recorded 3 embedded indications that were identified with the 60° shear during PSI - 2 indications were below 10/1990  1st Interval/                          recordable levels; 2                            2RC-01-BB/SGC-031                                    100%
(B2R02)    1st Period                      therefore, they were not required to be compared to code standards. The original Indication #1 from PSI was unacceptable based on IWB-3511.
Resize of 3 embedded 9/1993  1st Interval/                    indications recorded with 2                            2RC-01-BB/SGC-031                                    N/A (B2R04)    2nd Period                    60° shear during PSI - no change.
Recorded the same 97.5% -
11/1999  2nd Interval/                      embedded indications 2                            2RC-01-BB/SGC-031                              limited due to (B2R08)    1st Period                        identified in PSI and ID weld pads Geometry with no change.
Recorded the same 97.5% -
10/2008  3rd Interval/                      embedded indications 2                            2RC-01-BB/SGC-031                              limited due to (B2R14)    1st Period                        identified in PSI and ID weld pads Geometry with no change.
Recorded 6 embedded reflectors (3 identified in 10/1990  1st Interval/
2                            2RC-01-BB/SGC-05        PSI). Acceptable in          100%
(B2R02)    1st Period accordance with IWB-3511.
9/2002  2nd Interval/                      Recorded 3 embedded 2                            2RC-01-BB/SGC-05                                    100%
(B2R10)    2nd Period                    reflectors with no change.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 87 of 98) 4/2013      3rd Interval/
2                                  2RC-01-BB/SGC-05                      NRI                  100%
(B2R17)      3rd Period Resize of 2 embedded indications recorded with 45° shear during PSI with 10/1990      1st Interval/
2                                  2RC-01-BC/SGC-051          no significant changes.          N/A (B2R02)      1st Period Both indications were unacceptable based on IWB-3511 during PSI.
Resize of 2 embedded 9/1993      1st Interval/                          indications recorded with 2                                  2RC-01-BC/SGC-051                                            N/A (B2R04)      2nd Period                              45° shear during PSI with no changes.
90% Limited Recorded 5 embedded due to weld 1/1989      1st Interval/                          reflectors. 1 indication was 2                                  2RC-01-BD/SGC-061                                          pads and (B2R01)      1st Period                              unacceptable based on adjacent IWB-3511 during PSI nozzles Resize of 1 embedded indication previously 10/1993      1st Interval/
2                                  2RC-01-BD/SGC-061          recorded with 60° shear            N/A (B2R04)      2nd Period during PSI and B2R01 with no change 97% limited 10/2005    2nd Interval/                                                          due to level 2                                  2RC-01-BD/SGC-061                      NRI (B2R12)      3rd Period                                                          taps and weld pads 97% limited 9/2011      3rd Interval/                                                          due to level 2                                  2RC-01-BD/SGC-061                      NRI (B2R16)      2nd Period                                                            taps and weld pads Note:
: 1.      Welds were found to have embedded indications during the PSI examinations. Relief was requested to accept these flaws which was approved through an NRC safety evaluation on 10/29/1986. Successive examinations were performed the welds and the embedded indications were either found to have no change in sizing or were found to be non-recordable due to improved inspection methods during later ISI Intervals.
: 2.      Examination of the upper secondary shell - shell cone welds (1RC-01-BA/SGW-05, 1RC-01-BB/SGW-05, 1RC-01-BC/SGW-05, and 1RC-01-BD/SGW-05) were not examined during the Second or Third Intervals, due to the basis that these welds were not located at gross structural discontinuities as defined in ASME Section III NB-3213.2. The welds were removed from the Examination Category C-A Item Number C1.10 examination population as permitted by Table IWC-2500-1 Examination Category C-A Note 2.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 88 of 98)
Table B7. Byron SG Secondary Side Head Welds (C1.20)
Item Unit    Date    Interval/Period        Component ID    Exam Results      Coverage No.
C1.20          3/2005    2nd Interval/ 3rd 1                                      1RC-01-BA/SGW-08        NRI          100%
(B1R13)          Period 9/2012    3rd Interval/ 2nd 1                                      1RC-01-BA/SGW-08        NRI          100%
(B1R18)          Period 3/2020    4th Interval/ 2nd 1                                      1RC-01-BA/SGW-08        NRI          100%
(B1R23)          Period 3/1992    1st Interval/ 2nd 2                                      2RC-01-BA/SGC-08        NRI          100%
(B2R03)          Period 3/2004    2 Interval/ 2nd nd 2                                      2RC-01-BA/SGC-08        NRI          100%
(B2R11)          Period Recorded a spot indication with no through wall 4/2013      3rd Interval/ 3rd 2                                      2RC-01-BA/SGC-08    dimension.      100%
(B2R17)          Period Acceptable in accordance with IWC-3510
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 89 of 98)
Table B8. Byron SG Secondary Side Tubesheet-to-Shell Welds (C1.30)
Item                            Interval/
Unit        Date                        Component ID                    Exam Results          Coverage No.                              Period C1.30              3/2002    2nd Interval/                                                                100%
1                                    1RC-01-BA/SGW-02                        NRI2 (B1R11)      2nd Period 9/2009    3rd Interval/                                                                100%
1                                    1RC-01-BA/SGW-02                        NRI2 (B1R16)      1st Period 9/2018    4th Interval/                                                                100%
1                                    1RC-01-BA/SGW-02                        NRI2 (B1R22)      1st Period Recorded 4 embedded      90% - limited due 9/1990    1st Interval/                                  reflectors and 1 was  to weld pads and 2                                    2RC-01-BA/SGC-021 (B2R02)      1st Period                                    unacceptable based    adjacent nozzles on IWC-3510 Resize of 4 embedded            N/A 9/1993    1st Interval/
2                                    2RC-01-BA/SGC-02            indications identified in (B2R04)      2nd Period B2R02 - No change Resize of 4 embedded            N/A 8/1996    1st Interval/
2                                    2RC-01-BA/SGC-02            indications identified in (B2R06)      3rd Period B2R02 - No change Recorded 4 embedded        96.4% - limited 10/1999    2nd Interval/                                indications previously  due to weld pads 2                                    2RC-01-BA/SGC-02 (B2R08)      1st Period                                    identified in B2R02 -    and adjacent No change            nozzles Recorded 4 embedded        96.4% - limited 10/2008    3rd Interval/                                indications previously  due to weld pads 2                                    2RC-01-BA/SGC-02 (B2R14)      1st Period                                    identified in B2R02 -    and adjacent No change            nozzles Recorded ID Geometry        95.9% limited and 3 embedded        due to adjacent 4/2019    4th Interval/
2                                    2RC-01-BA/SGC-02                reflectors that were      nozzles (B2R21)      1st Period determined to be non-relevant.
1/1989    1st Interval/                                                                100%
2                                    2RC-01-BC/SGC-02                        NRI (B2R01)      1st Period 9/1993    1st Interval/                                                                100%
2                                    2RC-01-BC/SGC-02                        NRI (B2R04)      2nd Period 2/1989    1st Interval/                                                                100%
2                                    2RC-01-BD/SGC-02                        NRI (B2R01)      1st Period 10/1993    1st Interval/                                                                100%
2                                    2RC-01-BD/SGC-02                        NRI (B2R04)      2nd Period Note:
: 1.      Weld 2RC-01-BA/SGC-02 was found to have 4 embedded indications during the 1st ISI Interval examination.
Successive examinations were performed and the embedded indications were either found to be non-relevant or did not change in size during the latest 4th ISI Interval examination.
: 2.      Noted ID geometry.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 90 of 98)
CALVERT CLIFFS, UNITS 1 AND 2, INSPECTION HISTORY Currently, the SG components for Calvert Cliffs, Units 1 and 2, satisfy all of the inspection requirements of ASME Code, Section XI, 2013 Edition. A summary of the inspection history for Item Numbers B2.40, B3.130, C1.10, C1.20, and C1.30 for Calvert Cliffs, Units 1 and 2, is provided in Tables B9 through B13.
Table B9. Calvert Cliffs SG Primary Side Tubesheet-to-Shell Welds (B2.40)
Item                                                                        Exam Unit      Date        Interval/Period        Component ID                  Coverage No.                                                                      Results 3/2010 1                  3rd Interval/ 3rd Period    11-W22              NRI        97.7%
(1RFO19) 2/2014 1                4th Interval/ 2nd Period    11-W22              NRI        97.6%
(1RFO21)
B2.40 3/2007 2                  3rd Interval/ 3rd Period    21-W22              NRI        99.5%
(2RFO16) 2/2013 2                4th Interval/ 2nd Period    21-W22              NRI        97.0%
(2RFO19)
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 91 of 98)
Table B10. Calvert Cliffs SG Primary Side Nozzle-to-Vessel Welds (B3.130)
Item                                                                                      Exam Unit          Date          Interval/Period        Component ID                          Coverage No.                                                                                    Results B3.130                    2/2006 1                        3rd Interval/ 3rd Period    SG W5              NRI              84%
(1RFO17) 2/2014 1                        4th Interval/ 2nd Period      SG W5              NRI            81.6%
(1RFO21) 3/2010 1                        3rd Interval/ 3rd Period    SG W6              NRI            74.2%
(1RFO19) 2/2018 1                        4th Interval/ 3rd Period    SG W6              NRI            73.8%
(1RFO23) 3/2010 1                        4th Interval/ 1st Period    SG W7              NRI            74.2%
(1RFO19) 2/2006 1                        3rd Interval/ 3rd Period    SG W5              NRI              84%
(1RFO17) 2/2010 1                        4th Interval/ 1st Period    SG W5              NRI            84.3%
(1RFO19) 2/2010 1                        4th Interval/ 1st Period    SG W6              NRI            74.2%
(1RFO19) 2/2010 1                        4th Interval/ 1st Period    SG W7              NRI            74.2%
(1RFO19) 2/2013 2                        4th Interval/ 2nd Period      SG W5              NRI            80.2%
(2RFO19) 2/2013 2                        4th Interval/ 2nd Period      SG W6              NRI            73.8%
(2RFO19) 3/2011 2                        4th Interval/ 1st Period    SG W7              NRI            78.0%
(2RFO18) 3/2009 2                        3rd Interval/ 3rd Period    SG W5              NRI            100%
(2RFO17) 3/2013 2                        4th Interval/ 2nd Period      SG W5              NRI            80.2%
(2RFO19) 3/2009 2                        3rd Interval/ 3rd Period    SG W6              NRI            100%
(2RFO17) 3/2013 2                        4th Interval/ 2nd Period      SG W6              NRI            73.8%
(2RFO19) 3/2013 2                        4th Interval/ 2nd Period      SG W7              NRI            73.8%
(2RFO19)
Note:
: 1.      Since the steam generator replacement was performed during the first period of the 3 rd Interval, Paragraph IWB-2412(b) was utilized which only required 25% of the population to be inspected in each of the Second and Third Periods, for a total of 50% inspected by end of Third Period. This allowed Calvert Cliffs to forego the inspection of weld W7 for both Units.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 92 of 98)
Table B11. Calvert Cliffs SG Secondary Side Shell Welds (C1.10)
Item                                                                    Exam Unit      Date        Interval/Period      Component ID              Coverage No.                                                                  Results C1.10              3/2010 1                  3rd Interval/ 3rd Period    SG-11-4        NRI      93.1%
(1RFO19) 3/2018 1                  4th Interval/ 3rd Period    SG-11-4        NRI      100%
(1RFO23) 3/2005 2                  3rd Interval/ 2nd Period    SG-22-4        NRI      99.0%
(2RFO15) 2/2011 2                  4th Interval/ 1st Period    SG-22-4        NRI      99.0%
(2RFO18)
Table B12. Calvert Cliffs SG Secondary Side Head Welds (C1.20)
Item                                                                    Exam Unit      Date        Interval/Period      Component ID              Coverage No.                                                                  Results C1.20              3/2010 1                  3rd Interval/ 3rd Period    SG-11-3        NRI      100%
(1RFO19)
Unchanged 3/2018                                                ID 1                  4th Interval/ 3rd Period    SG-11-3                  100%
(1RFO23)                                            Geometry Indication Base 3/2009                                              material 2                  3rd Interval/ 3rd Period    SG-21-3                  100%
(2RFO17)                                              laminar indication 2/2019 2                  4th Interval/ 3rd Period    SG-21-3        NRI      100%
(2RFO22)
Table B13. Calvert Cliffs SG Secondary Side Tubesheet-to-Shell Welds (C1.30)
Item                                                                    Exam Unit      Date        Interval/Period      Component ID              Coverage No.                                                                  Results C1.30              3/2010 1                  3rd Interval/ 3rd Period  SG W65    NRI      98.6%
(1RFO19) 3/2018 1                  4th Interval/ 3rd Period  SG W65    NRI      100%
(1RFO23) 3/2007 2                  3rd Interval/ 3rd Period  SG W65      NRI      100%
(2RFO16) 2/2017 2                  4th Interval/ 3rd Period  SG W65      NRI      100%
(2RFO21)
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 93 of 98)
GINNA INSPECTION HISTORY Currently, the SG components for Ginna satisfy all of the inspection requirements of ASME Code, Section XI, 2013 Edition. A summary of the inspection history for Item Numbers B2.40, C1.10, C1.20, and C1.30 for Ginna is provided in Tables B14 through B17.
Table B14. Ginna SG Primary Side Tubesheet-to-Shell Welds (B2.40)
Item                                                                        Exam Unit      Date        Interval/Period      Component ID                Coverage No.                                                                      Results B2.40                4/2008 1                4th Interval/ 3rd Period  LHTSW-AR          NRI        99%
(O2008) 10/2015 1                5th Interval/ 2nd Period  LHTSW-AR          NRI      96.7%
(O2015)
Table B15. Ginna SG Secondary Side Shell Welds (C1.10)
Item                                                                        Exam Unit      Date        Interval/Period      Component ID                Coverage No.                                                                      Results C1.10                9/2009 1                4th Interval/ 3rd Period    TUS-BR            NRI        98%
(O2009) 5/2017 1                5th Interval/ 3rd Period    TUS-BR            NRI      99.2%
(O2017) 1/2002 1                4th Interval/ 1st Period    LST-AR            NRI      >90%
(O2002) 11/2012 1                5th Interval/ 1st Period    LST-AR            NRI      100%
(O2012)
Table B16. Ginna SG Secondary Side Head Welds (C1.20)
Item                                                                        Exam Unit      Date        Interval/Period      Component ID                Coverage No.                                                                      Results C1.20                1/2002 1                4th Interval/ 1st Period    USH-AR            NRI      >90%
(O2002) 10/2012 1                5th Interval/ 1st Period    USH-AR            NRI      99.7%
(O2012)
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 94 of 98)
Table B17. Ginna SG Secondary Side Tubesheet-to-Shell Welds (C1.30)
Item                                                                  Exam Unit      Date        Interval/Period      Component ID              Coverage No.                                                                Results C1.30            10/2015 1                5th Interval/ 2nd Period  TSSW-AR        NRI      99.49%
(O2015) 10/2006 1                4th Interval/ 2nd Period  TSSW-BR        NRI        99.6%
(O2006)
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 95 of 98)
APPENDIX C RESULTS OF INDUSTRY SURVEY
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 96 of 98)
Overall Industry Inspection Summary for Code Item Numbers B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30 The results of an industry survey of past inspections of SG nozzle-to-shell welds, inside radius sections and shell welds are summarized in Reference [1]. Table C1 provides a summary of the combined survey results for Item Numbers B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30. The results of the industry survey identified numerous steam generator (SG) examinations being performed with no service-induced flaws being detected. Performing these examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international boiling water reactor (BWR) and pressurized water reactor (PWR) units responded to the survey and provided information representing all PWR plant designs currently in operation in the United States. This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR nuclear steam supply system (NSSS) vendors (i.e., Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse). A total of 1,324 examinations for the components of the affected Item Numbers were conducted, with 1,098 of these specifically for PWR components. The majority of the PWR examinations were performed on SG welds.
A relatively small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service-induced. For Item Number B2.40, examinations at two units at a single plant site identified multiple flaws exceeding the acceptance criteria of ASME Code Section XI; however, these were determined to be subsurface-embedded fabrication flaws and not service-induced (see Note 1 for Table 1). For Item Number C1.20, two PWR units reported flaws exceeding the acceptance criteria of ASME Code, Section XI. In the first unit, a single flaw was identified, and was evaluated as an inner diameter surface imperfection. Reference [2] indicates that this was a spot indication with no measurable through-wall depth. This indication is therefore not considered to be service-induced but rather fabrication-related. A flaw evaluation per IWC-3600 was performed for this flaw and it was found to be acceptable for continued operation. In the second unit, multiple flaws were identified (see Note 2 for Table 1). As discussed in References [3] and [4], these flaws were most likely subsurface weld defects typical of thick vessel welds and not service- induced.
A flaw evaluation per IWC-3600 was performed for these flaws and they were found to be acceptable for continued operation.
 
==References:==
: 1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906, ADAMS Accession No. ML20225A141.
: 2. Letter from F. A. Kearney (Exelon) to U.S. NRC, Byron Station Unit 2 90-Day lnservice Inspection Report for Interval 3, Period 3, (B2R17), dated July 29, 2013, Docket No. 50-455, ADAMS Accession Number ML13217A093.
: 3. Letter from J. M. Sorensen (NMC) to U.S. NRC, Unit 1 lnservice Inspection Summary Report,
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 97 of 98)
Interval 3, Period 3 Refueling Outage Dates 1-19-2001 to 2-25-2001 Cycle 20 / 05-26-99 to 02-25-2001, May 29, 2001, Docket Nos. 50-282 and 50-306, ADAMS Accession Number ML011550346.
: 4. Letter from J. P. Solymossy (NMC) to U. S. NRC, Response to Opportunity for Comment on Task Interface Agreement (TIA) 2003-01, Application of ASME Code Section XI, IWB-2430 Requirements Associated with Scope of Volumetric Weld Expansion at the Prairie Island Nuclear Generating Plant (Tac Nos. MB7294 and MB7295), April 4, 2003, Docket Nos. 50-282 and 50-306, ADAMS Accession Number ML031040553.
 
10 CFR 50.55a Proposed Alternative I4R-17 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-23 for Byron Station, Units 1 and 2, Proposed Alternative ISI-05-018 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and Proposed Alternative I6R-10 for Ginna Revision 1 (Page 98 of 98)
Table C1. Summary of Survey Results Number of Examinations                        Number of Reportable Indications Item Number BWR          PWR            Total          BWR            PWR                Total B2.31            0          30                30              0            0                    0 B2.32            0          13                13              0            0                    0 B2.40            0          183              183              0          Note 1              Note 1 B3.130            0          135              135              0            0                    0 C1.10          140          305              445              0            0                    0 C1.20            54          319              373              0          Note 2              Note 2 C1.30            32          113              145              0            0                    0 Totals          226        1,098            1,324              0            ---                  ---
Note:
: 1.      Two PWR W-2 Loop units at a single plant reported multiple subsurface embedded fabrication flaws.
: 2.      A single PWR W-2 Loop unit reported multiple flaws [3, 4 ].}}

Revision as of 10:38, 13 November 2024

Proposed Alternative for Examinations of Examination Categories B-B, B-D, and C-A Steam Generator Pressure Retaining Welds and Full Penetration Welded Nozzles
ML23284A259
Person / Time
Site: Calvert Cliffs, Byron, Braidwood, Ginna  Constellation icon.png
Issue date: 10/11/2023
From: David Gudger
Constellation Energy Generation
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-23-108
Download: ML23284A259 (1)
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