View source for RS-16-154, Relief Request I3R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval

{{Adams
| number = ML16203A081
| issue date = 07/21/2016
| title = Relief Request I3R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval
| author name = Gullott D
| author affiliation = Exelon Generation Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000456, 05000457
| license number = NPF-072, NPF-077
| contact person = 
| case reference number = RS-16-154
| document type = Letter type:RS
| page count = 14
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:ExeLon G 4300 Winfield Road Warrenville, IL 60555 www.exeloncorp.com RS-16-154                                                                                10 CFR 50.55a July 21, 2016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NFP-77 NRC Docket Nos. STN-50-456 and STN 50-457

==Subject:==
Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval

==References:==
: 1) WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011
: 2) Letter from R. A. Nelson (NRC) to W. A. Nowinowski (PWROG), "Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation Regarding Pressurized Water Reactor Owners Group Topical Report WCAP-16168-NP-A, Revision 2, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,"' dated July 26, 2011 In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC, (EGC) hereby requests NRC approval of the attached relief request associated with the Third 10-year Inservice Inspection (ISI) Program Interval for Braidwood Station, Units 1 and 2. The third interval of the Braidwood Station ISI program complies with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition through the 2003 Addenda. The Braidwood Station Third 10-year ISI intervals are currently scheduled to end on July 28, 2018 for Unit 1 and October 16, 2018 for Unit 2.
NRC approval of attached Relief Request 13R-17 is requested based on the justification provided in topical report WCAP-16168-NP-A, Revision 3 (Reference 1). This topical report demonstrated that extending the ASME B&PV Code, Section XI ISI interval from the current 10 years to 20 years for reactor pressure vessel (RPV) pressure containing welds is an alternative inspection interval that provides an acceptable level of quality and safety. The NRC approved WCAP-16168 in Reference 2. Relief Request 13R-17 specifically addresses Examination Categories B-A and B-D applicable to the RPV for Braidwood Station, Units 1 and 2.

July 21, 2016 U. S. Nuclear Regulatory Commission Page 2 EGC requests approval of this relief request by July 21, 2017 to support the Braidwood Station, Unit 1 refueling outage (A1 R20) scheduled to commence on April 9, 2018.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Joseph A. Bauer at (630) 657-2804.
Respectfully, David M. Gullott Manager  Licensing Exelon Generation Company, LLC

==Attachment:==
10 CFR 50.55a Relief Request 13R-17 CC'     NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, Braidwood Station NRR Project Manager, Braidwood Station Ilinois 'Emergency Management Agency  Division of Nuclear Safety

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 1 of 12

===1.0 ASME Code Components Affected===
The affected components are the Braidwood Station, Unit 1 and Unit 2 Reactor Pressure Vessel (RPV); specifically, the following American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel (B&PV) Code, Section XI (Reference 1) examination categories and item numbers covering examinations of the RPV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME B&PV Code, Section XI.
Category B-A welds are defined as "Pressure Retaining Welds in Reactor Vessel." Category B-D welds are defined as "Full Penetration Welded Nozzles in Vessels" Examination Category          Item No. Description B-A               B1.11        Circumferential Shell Welds B-A               131.21       Circumferential Head Welds B-A               B1.30        Shell-to-Flange Weld B-D               B3.90        Nozzle-to-Vessel Welds B-D               B3.100       Nozzle Inside Radius Section 1   * *
* IM
             * * - * *
* M * - *.
The Third Interval Inservice Inspection (ISI) program is based on the ASME B&PV Code, Section XI, 2001 Edition through 2003 Addenda (Reference 1). Throughout this request the above examination categories are referred to as "the subject examinations' and the ASME B&PV Code. Section XI, is referred to as "the Code."
The applicable Code for the subsequent (i.e., Fourth) ISI interval will be implemented in accordance with the requirements of 10 CFR 50.55a.

===3.0 Applicable Code Requirement===
ASME Section XI Paragraph IWB-2412, "Inspection Program B," requires volumetric and surface examination of essentially 100% of RPV pressure retaining welds identified in Table IWB-2500-1 once each 10-year interval. The Braidwood Station Third 10-year ISI intervals are currently scheduled to end on July 28, 2018 for Unit 1 and October 16, 2018 for Unit 2.

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 2 of 12

===4.0 Reason for Request===
Pursuant to 10 CFR 50.55a(z)(1), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety. An alternative is requested from the requirement of IWB-2412, "Inspection Program B," that volumetric examinations of essentially 100% of RPV pressure retaining, Examination Category B-A and B-D items, be performed once each 10-year interval.
Extension of the interval between examinations of Category B-A and B-D items from 10 years to a maximum of 20 years will result in a reduction in person-rem exposure and examination costs.
Exelon Generation Company, LLC (EGC) proposes to defer the ASME B&PV Code required volumetric examinations of specific Braidwood Station Unit 1 and Unit 2 RPV pressure retaining Examination Category B-A and B-D items. These examinations are currently scheduled to be performed during the Third ISI Interval in 2018. Upon approval of this relief request, these required examinations would be performed during the Fourth ISI Interval for each unit in 2027 plus or minus one refueling outage. These dates are consistent with those provided in Pressurized Water Reactor Owners Group (PWROG) letter OG-06-356 (Reference
: 10) and the latest implementation plan provided in PWROG letter OG-10-238 (Reference 2).
Tables 1-1 and 1-2 list the applicable examination areas addressed under this relief request for Braidwood Station, Unit 1 and Unit 2 respectively.
In accordance with 10 CFR 50.55a(z)(1), an alternate inspection interval is requested on the basis that the current interval can be extended based on a negligible change in risk when compared to the risk criteria specified in Regulatory Guide 1.174 (Reference 3).
The methodology used to demonstrate the acceptability of extending the inspection intervals for Category B-A and B-D items, based on a negligible change in risk, is contained in WCAP-16168-NP-A, Revision 3, (Reference 4). This methodology was used to develop a pilot plant analysis for RPVs in Westinghouse, Combustion Engineering, and Babcock and Wilcox plant designs, and is an extension of the work that was performed as part of the NRC Pressurized Thermal Shock (PTS) Risk Study (Reference 5). The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant specific basis, as identified in WCAP-16168-NP-A, Revision 3, are shown in Table 2. Table 2 lists the critical parameters investigated in WCAP-16168-NP-A, Revision 3 and compares the results of the Westinghouse pilot plant to those of Braidwood Units 1 and 2 for an ISI Interval extension.
Tables 3-1, 3-2, 4-1, and 4-2 provide additional information required by the NRC as noted in Appendix A of WCAP-16168-NP-A, Revision 3.
By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology for Braidwood Station, Units 1 and 2 is determined to be acceptable.

ATTACHRflENT 10 CHI,                RELIEF REQUEST I3R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inserviee Inspection Interval lug Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 3of12 Table 1-1 Examination Category, Item Number, Component Description and Component Identification of
_                Applicable Examinations for Braidwood Unit I Exa1m      -  Exam Item                        Component Description                                            BSI Weld Number Cate9 o ry      Number B-A           131.11      _Reactor Vessel Shell-to-Reactor Vessel Shell                                    '    1 RV-01-003 Reactor Vessel Shell-to-Reactor Vessel Shell                                          11RV-01-004 Reactor Vessel Shell-to-Dutc                                                          1 RV-02-002 131.21   1 Top Head--to-Upper Center Disc                                                         1 RV-03-002 131.30      Reactor Vessel-to-Flange                                                              11RV-01-005 B-D           B3.90       Outlet Nozzle @ 22 Degrees                                                            1RV-01-006 Inlet Nozzle @ 67 Degrees                                                             " ITV-01-007 Inlet Nozzle @ 113 Degrees                                                       !    1RV-01-008 Outlet Nozzle @ 158 Degrees                                                      I    1 RV-01-009 C  Outlet Nozzle @ 202 Degrees                                                           1 RV-01-010 j lnlet Nozzle C,   247 Degrees                                                         1 RV-01-C11 I  Inset Nozzle @ 293 Degrees                           _                                11RV-01-C12 Outlet Nozzle @ 338 Degrees                                                           i RV-01-013 B3.100      Outlet Nozzle Inner Radius @ 22 Degrees                                          i    11RV-01-014 lnlet Nozzle Inner Radius @ 67 De rees                                                1 RV-01-015 ln`et Nozzle --Inner Radius          @ 1 `3 Decrees
                                                    ....._.....---..._...._....---...                             1RV-01-016 Outlet Nozz!e Inner Radius (@ 158 Degrees                                               RV-01-017 Outlet Nozzle Inner Radius                      202 Degrees                             13V-01-018 Inlet Nozzle Inner Radius @ 247 Degrees                                               1 `:RV-01-019 In"et Nozzle Inner Radius @ 293 Decrees                                               " RV-01-02C Outlet Nozzle Inner Rad&deg;us ,@ 338 Degrees                                               W-01-021

ATTACHMENT
                            '1 0 CFR 50.55a RELIEF REQUEST MKI-1 7 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 11=- *f~ ~r~
Table 1-2 Examination Category, Item Number, Component Description and Component Identification of Applicable Examinations for Braidwood Unit 2 Exam ---' Exam Item                         Component Description                                                                                  F_ ISI Weld Number Category      Number B-A          131.1 1      Reactor Vessel Shell-to-Reactor Vessel Shell                                                                                   2RV-01 -003 Reactor Vessel Shell-to-Reactor Vessel Shell                                                                                   2RV-01 -004 Reactor Vessel Shell-to-Dut0man                                                                                                2 R V 002 B1.21       To Head-to-Upper Center Disc                                                                                                    2RV-03-002 81.30   I Reactor Vessel-to-Flange                                                                                                          2RV-01 -005 B-D          83.90   I Outlet Nozzle @ 22 Degrees                                                                                                        2 RV-01 -006 Inset Nozzle @ 67 Degrees                                                                                                       2RV-01-007 Inset Nozzle @ ", A 3 Degrees                                                                                                   2RV-01-008 Outlet Nozzle @ 158 Degrees                                                                                                    2RV-01-009 Outlet Nozzle @ 202 Degrees                                                                                                    2 RV-01 -0 10 In.-et Nozzle @ 247 Degrees                                                                                                     2 RV-01 -0 11
_Inset Nozzle         293                                                                                                                         1 -0 1 2 Ou"let Nozzle @ 338 Decrees                                                                                                    2RV-0'f. -013 33.100 Outlet Nozzle Inner Radius                            22 Degrees P                                                                           2RV-0f-014 Inset Nozzle InnerRacius @ 67 Degrees                                                                                           2RV-0f-015 ln`et Nozzle Inner Radius @ 1'L 3 Deg. rees            . . . . . . . . . . . . . . . . . . . . - --- - -------------------           -. . . . . . ..
2RV0",-0"L6 Outlet Nozzle Inner Rad us @ 158 Degrees                                                                                        2RV-01 -0" 7 Outlet Nozzle Inner
                       ...........                       Rad'us
                                                ....................... ' @ 202 Degrees                                                                    2RV-01-0"8 ln'Let Nozzle Inner .Radius @ 247 Degrees                                                                                       2RV-01 -01. 9 11.1..'et Nozzle Inner Radius @ 293 Degrees                                                                                     2!RV-0', -020 Outlet Nozzle Inner Radius @, 338 De, -ees                                                                                     2 RV-0" -02f

10 CFRI 50.55a REUEF FIE0UEST l3R-I 7 Request for ReHefr for Aiternadve RegWremants for Reactor Pressure !Vessel 6nservice lnspecttion hteirvai in Accordance @fflith 10 CFR 50.55a(z)(1)
Revs sion
                                                'age 5 of 12 Table 2 Critical Parameters for the Application of Bounding Analysis for Braidwood Units 1 and 2 Parameter                  Pilot Plant Basis      Plant-Specific Basis    :additional Evaluation Required Dominant Pressurized             NRC PTS Risk Study       PTS Generalization            No Therma', Shock (PTS)             (Reference 7)            Study (Reference 6)
Transients in the NRC PTS Risk Study are Applicable I' Through-Wall Cracking            1.76E-08 Events          Unit 1: 2.09E-16 Events       No Frequency (TWCF)                 per year                 per year (caculated per (Reference 4)            Reference 4)
Unit 2: 2.04'E-15 Events per year (caaculated per I                          Reference 4)
Frequency and Severity of        7 ' eatup / cooldown     Sounded by 7                  No Design Basis Transients          cycles per year          )eatup / cooldown (Reference 4)            cycles per year Cadding Layers                   S.ncle Layer             S;r:Cle Layer No (S rgie/Ulultiple)               (Reference 4)

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 6 of 12 Tables 3-1 and 3-2 provide a summary of the latest reactor vessel inspections for Braidwood Station, Units 1 and 2 respectively, and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed for both Braidwood reactor vessels.
Table 3-1 Additional Information Pertaining to Reactor Vessel Inspection for Braidwood Unit 1 Inspection methodology:         The latest ISI was conducted in accordance with the ASME Code, Section XI and Section V, 1989 Edition, with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Examinations of Category B-A and B-D welds were performed to ASME Section XI, Appendix VIII, 1989 Edition with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future inservice inspections will be performed to ASME Section XI, Appendix VIII requirements.
Number of past inspections:     Two 10-year inservice inspections have been performed.
Number of indications found:    There were three indications identified in the beltline region during the most recently completed inservice inspection. One subsurface indication is located in the nozzle shell to intermediate shell circumferential weld seam (Item 4 in Table 3-1) and two subsurface indications are located in the intermediate shell to lower shell circumferential weld seam (Item 5 in Table 3-1). All three indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. None of these indications are within the inner 1110"' or 1 inch of the reactor vessel thickness; therefore, they are inherently acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a (Reference 7).
Proposed inspection             The third inservice inspection currently is scheduled for 2018. This inspection schedule for balance of         will be performed in 2027 plus or minus one refueling outage. The plant life:                     proposed inspection date for Braidwood Unit 1 is consistent with the latest implementation plan presented in OG-10-238 (Reference 2).

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 7 of 12 Table 3-2 Additional Information Pertaining to Reactor Vessel Inspection for Braidwood Unit 2 Inspection methodology:        The latest ISI was conducted in accordance with the ASME Code, Section XI and Section V, 1989 Edition, with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Examinations of Category B-A and B-D welds were performed to ASME Section XI, Appendix VIII, 1989 Edition with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi).
Future inservice inspections will be performed to ASME Section XI, Appendix VIII requirements.
Number of past inspections: Two 10-year inservice inspections have been performed.
Number of indications          There were two indications identified in the beltline region during the most found:                         recently completed inservice inspection. One subsurface indication is located in the nozzle shell to intermediate shell circumferential weld seam (Item 4 in Table 3-2) and another subsurface indication is located in the intermediate shell to lower shell circumferential weld seam (Item 5 in Table 3-2). Both indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. One indication is within the inner 1/10t&deg;'
or 1" of the reactor vessel thickness. This indication is acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7), since the number of flaws is less than the allowable number of flaws for each flaw size increment. A disposition of this flaw against the limits of the Alternate PTS Rule is shown in the table below. The following indication is located within the forging material of the reactor vessel beltline.
Through-Wall                                                        Scaled Extent, TWE                                                                                                                           Number of maximum forging flaws TWEMIN                         TWEr\AAX                         number of (Axial/Circ.)
forging flaws 0.075                               0.375                                     76                                              +                             1 (0/1) 0.125                              0.375                                      30                                                                            1 (0/1) 0.175                              0.375                                         8                                                                          0 .(0/0)
Proposed inspection            The third inservice inspection currently is scheduled for 2018. This schedule for balance of        inspection will be performed in 2027 plus or minus one refueling plant life:                    outage. The proposed inspection date for Braidwood Station, Unit 2 is consistent with the latest implementation plan presented in OG-10-238 (Reference 2).

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 8 of 12 Tables 4-1 and 4-2 summarize the inputs and outputs for the calculation of through-wall cracking frequency (TWCF) for Braidwood Station, Units 1 and 2 respectively.
Table 4-1:
Details of TWCF Calculation for Braidwood Unit 1 at 57 Effective Full-Power Years (EFPY)
Inputs Reactor Coolant System Temperature, TRcs [&deg;F]:                    N/A                             Twall [inches]:                                  8.625 No.              Region and                        Material Heat No.
Component Cum           Ni(1)    R.G.0)           CF0)       RTNDT(u)(1)         Fluence [10 y n/cmz,
[wt%]         [wt%]     1.99              [&deg;F]           [&deg;F]
Description                                                                                                                             E>1.0 MeV (1)
Pos.                                                          ]
1      Nozzle Shell (NS) Forging                   5P-7016                    0.04          0.73      1.1             26             10                      1.14 2      Intermediate Shell (IS) Forging        [49D383/49C344]-1-1             0.05          0.73      1.1             31            -30                      3.19 3      Lower Shell (LS) Forging               [49D867/49C813]-1-1             0.05          0.74      1.1             31            -20                      3.19 4      NS to IS Circ. Weld Seam                     H4498                     0.04          0.46      1.1             54            -25                      1.14 5      IS to LS Circ. Weld Seam                     442011                    0.03          0.67      1.1             41             40                      3.06 Outputs Methodology Used to Calculate AT30:
Regulatory Guide 1.99, Revision 2(2)
Controlling Material                             Fluence                   FF Region No. (From Above)      RTMAx-xx [IR]            19      Z           (Fluence        AT;Q [&deg;F]       TWCF95-XX
[10 n/cm ,
Factor)
E > 1.0 MeV]
Limiting Forging  FO                                    1                496.62                  1.14                1.037            26.95            8.38E-17 Limiting Circumferential Weld - CW                               5                552.78                  3.06                1.295            53.11           0.00E+00 TWC F95-TOTAL(aFoTWC FQ;-Fo +acwTWCF95-cw).                                                 2.09E-16 (1) Reference 8 (2) Reference 9

rl VAC rH~ V-
                                                                                       - iEO'f 10 CFR b0.5ba RELIEF REQUEST I3R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 9 of 12 Table 4-2:
Details of TWCF Calculation for Braidwood Unit 2 at 57 Effective Full-Power Years (EFPY)
Inputs Reactor Coolant System Temperature, TRCs [&deg;F]                                   N/A                            TWa [inches]:                                8.625 No.        Region and Component                     Material Heat No.
Cu(1)          Ni(1)    R.G.(1)      CF(1)      RTNpT()(1) [&deg;F]        Fluence [10 V n/cm ,
Description
[ wt% ]       ['srt% ]    1.99          [OF]                                E> 1.0 141eV] (1)
Pos.
1      Nozzle Shell (NS) Forging                      5P-7056                      0.04          0.90      1.1          26               30                        1.11 2      Intermediate Shell (IS) Forging          [49D963/49C904]-1-1                0.03          0.71      1.1          20              -30                        3.16 3      Lower Shell (LS) Forging                   [50D102/50C97]-1-1               0.06          0.76      1.1          37              -30                        3.16 4      NS to IS Circ. Weld Seam                        H4498                       0.04          0.46      1.1          54              -25                        1.11 5      IS to LS Circ. Weld Seam                       442011                       0.03          0.67      1.1          41               40                        3.03 Outputs Methodology Used to Calculate AT3fl:
Regulatory M      Guide 1.99. Revision 2(2)
Controlling Material                                 Fluence                    FF Region No, (From Above)       RT.,, _   [&deg;R]                                   (Fluence       AT;:;[&deg;F]         TWCF :_
[10 j} n/cm2, Factor)
E > 1.0 MeV]
Limiting Forging  FO                                   1                  516.43                     1.11                 1.029             26.76           8.18E-16 Limiting Circumferential Weld - CW                             5                  552.69                     3.03              1  1.293             53.02           0.00E+00 TWCF95-T0TAL(ar0TWCF,_~0 + a, .,TWCF95_cw):                                              2.04E-15 (1) Reference 8 (2) Reference 9

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 10 of 12 6.0 Duration of Progosed Alternative This request is applicable to the Braidwood Station, Units 1 and 2 Inservice Inspection program for the third and fourth 10-year inspection intervals.
7.0 Precedents
: 1.  "Millstone Power Station, Unit No. 3  Alternative Request IR-3-27 for Implementation of Extended Reactor Vessel Inservice Inspection Interval (CAC NO. MF5868)," dated February 16, 2016 (ADAMS Accession Number ML16038A001)
: 2.  "Joseph M. Farley, Unit 1. Alternative to Inservice Inspection (CAC No. MF6475)," dated February 1, 2016 (ADAMS Accession Number ML16013A348)
: 3.  "Diablo Canyon Power Plant, Unit No. 1 - Request for Alternative RPV-U 1 -Extension to Allow Use of Alternate Reactor Inspection Interval Requirements (TAC No. MF4678),"
dated June 19, 2015 (ADAMS Accession Number ML15168A024)
: 4.  "Callaway Plant, Unit 1  Request for Relief 13R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAC No. MF3876)," dated February 10, 2015 (ADAMS Accession Number ML15035A148)
: 5.  "Shearon Harris Nuclear Power Plant, Unit 1  Relief from the Requirements of the ASME Code, Section XI (TAC NO. MF4113)," dated January 5, 2015 (ADAMS Accession Number ML14353A324)
: 6.  "Byron Station. Unit No. 1  Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596)," dated December 10.
2014 (ADAMS Accession Number ML14303A506, correction letter ML13113A016)
: 7.  "Wolf Creek Generating Station  Request for Relief Nos. 13R-08 and 13R-09 for the Third 10-Year Inservice Inspection Program Interval (TAC Nos. MF3321 and MF3322)," dated December 10, 2014 (ADAMS Accession Number ML14321A864)
: 8.  "Watts Bar Nuclear Plant, Unit 1- Request for Alternative 13-ISI-01 to Extend the Second Reactor Vessel Weld Inservice Inspection Interval (TAC No. MF2956)," dated November 28, 2014 (ADAMS Accession Number ML14314A987)
: 9.  "Sequoyah Nuclear Plant, Units 1 and 2  Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval (TAC Nos. MF2900 and MF2901)," dated August 1, 2014 (ADAMS Accession Number ML1418813920)
: 10. "Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAC Nos. MF1922 and MF1923),"
dated March 26, 2014 (ADAMS Accession Number ML14079A546)

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 11 of 12
: 11. "Vogtle Electric Generating Plant, Units 1 and 2  Request for Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT-06 (TAC Nos. MF2596 and MF2597)," dated March 20, 2014 (ADAMS Accession Number ML14030A570)
: 12. "Indian Point Nuclear Generating Unit No. 3  Safety Evaluation for Relief Request IP3-ISI-RR-06 for Reactor Vesel [sic] Weld Examinations (TAC No. MF3345)," dated July 23, 2014 (ADAMS Accession Number ML14198A331)
: 13. "Surry Power Station Units 1 and 2  Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, 2013 (ADAMS Accession Number ML13106A140)
: 14. "McGuire Nuclear Station, Unit 2, Relief 10-MN-002 to Extend the Inservice Inspection Interval for Reactor Vessel Category B-A and B-D Welds (TAC Nos. ME7329 and ME 7330)," dated September 6, 2012 (ADAMS Accession Number ML12249A175)
: 15. "Arkansas Nuclear One, Unit 2  Request for Alternative ANO2-ISI-004, to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations (TAC No. ME2508)," dated September 21, 2010 (ADAMS Accession Number ML102450654)
: 16. "Three Mile Island Nuclear Station, Unit 1 (TMI-1)  Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations, Proposed Alternative Request Nos. RR-09-01 and RR-09-02 (TAC Nos. ME2483 and ME2484)," dated September 21, 2010 (ADAMS Accession Number ML102390018)
: 17. "Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2)  Relief Request for Extension of the Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME3010)," dated July 12, 2010 (ADAMS Accession Number MI-101750402)
: 18. "Palo Verde Nuclear Generating Station, Units 1, 2, and 3  Relief Request No. RR-40, Reactor Vessel Weld Examination Interval Extension (TAC Nos. ME1634, ME1635, and ME1636)," dated February 22, 2010 (ADAMS Accession Number MI-100290415)
: 19. "Safety Evaluation of Relief Requests to Extend the Inservice Inspection Interval for Reactor Vessel Examinations for Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos. ME1478, ME1479, ME1480 and ME1481)," dated February 22, 2010 (ADAMS Accession Number ML100491550)

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 Request for Relief for Alternative Requirements for Reactor Pressure Vessel Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)
Revision 0 Page 12 of 12 8.0 References
: 1. ASME Boiler and Pressure Vessel Code, Section XI, 2001 Edition through 2003 Addenda, ASME International
: 2. PWROG Letter OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." PA-MSC-0120," July 12, 2010 (ADAMS Accession Number ML11153A033)
: 3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,' U.S. Nuclear Regulatory Commission, November 2002
: 4. Westinghouse Report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," October 2011 (ADAMS Accession Number ML113060207)
: 5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," U.S. Nuclear Regulatory Commission, March 2010
: 6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482)
: 7. 10 CFR Part 50.61 a, `Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75. No. 1, dated January 4, 2010 and No.
22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010
: 8. Westinghouse Report WCAP-17607-NP, Revision 0, "Braidwood Station Units 1 and 2 Reactor Vessel Integrity Evaluation to Support License Renewal Time-Limited Aging Analysis," December 2012 9,  NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988
: 10. OG-06-356, "Plan for Plant-Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, `Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval,' MUHP 5097-99, Task 2059," October 31, 2006}}