RS-17-031, Correction to Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval

Correction to Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval

ML17046A225
Person / Time
Site:	Braidwood   (NPF-072, NPF-077)
Issue date:	02/15/2017
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-17-031
Download: ML17046A225 (14)

Text

4,M WIIIfICld RC)aCI W7IICI)vllle, It. 60555 ExeLon

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Generation G30 657 l°°° Office RS-17-031

10 CFR 50.55a February 15, 2017 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:

Correction to Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval

References:

1) Letter from D. M. Gullott (Exelon Generation Company, LLC) to NRC, "Relief Request 13R-17 to Extend the Reactor Pressure Vessel Inservice Inspection Interval," dated July 21, 2016

2) WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," dated October 2011

3) 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 Reference 1, Exelon Generation Company, LLC, (EGC) requested NRC approval of Relief Request 13R-17 in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1).

Relief Request 13R-17 is associated with the Third 10-year Inservice Inspection (ISI) Program nterval 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.

As noted in Reference 1, NRC approval of Relief Request 13R-17 is requested based on the Justification provided in topical report WCAP-16168-NP-A, Revision 3 (Reference 2). This M

4.

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

February 15, 2017 U. S. Nuclear Regulatory Commission Page 2 NRC approved VVCAP-16168 in Reference 3. Relief Request 13R-17 specifically addresses Examination Categories B-A and B-D applicable to the RPV for Braidwood Station, Units 1 and 2.

During review of the planned Unit 1 Spring 2018 (Al R20) outage scope, an error was identified in Table Vl (i.e., list of Unit 1 subject welds) and Table 1-2 (i.e., list of Unit 2 subject welds) of the proposed relief request. Both the upper and the lower RPV heads have a Category B-A, Item 131.21 weld which falls under the scope of WCAP-16168-NP-A, Revision 3. Both of these upper welds (i.e., 1(2)RV-03-002) and lower welds (i.e., 1(2)RV-02-001) were listed on the noted tables in an early draft revision of the relief request. During review of the draft relief request, prior to submittal, it was concluded that the upper welds should be deleted from the tables since these welds were last inspected early in the second ISI interval. Extending the next inspection of the upper welds to near the end of the fourth ISI interval, in accordance with the subject relief request, would result in an inspection interval greater than the 20-year maximum interval approved in Reference 3. When deleting the subject welds from the tables, the lower welds vice upper welds were deleted in error.

This error has been corrected in Relief Request 13R-17, Revision 1 (attached). Specifically, the Lipper head Category B-A, Item B1.21 welds on Table 1-1 and Table 1-2 have been deleted; and the lower head Category B-A, Item B1.21 welds (i.e., 1(2)RV-02-001, "Dutchman-to-Lower Center Disc") have been added to Table 1-1 and Table 1-2, respectively. This change, denoted by a revision bar on Table 1-1 and Table 1-2, is the only change in Revision 1 of the subject relief request and has no impact on the technical justification for this relief request since, as noted above, these lower RPV head welds are within the scope of WCAP-16168-NP-A, Revision 3.

As noted in Reference 1, EGC requested approval of this relief request by July 21, 2017 to support the Braidwood Station, Unit 1 refueling outage (Al R20) scheduled to commence on April 9, 2018.

There are no regulatory commitments contained in this fetter. Should you have any questions concerning this letter, please contact Joseph A. Bauer at (630) 657-2804.

Respectfuly, j

David M. GLII:Ott Vanager Licensing Exelon Generation Company, =LC

Attachment:

`C CF R 50.55a Relief Request 3R-17, Revision cc:

NRC ;Regional Administrator. Region III NRC Senior Residert inspector, Braidwood Station NRR Project Veanacer, Braidwood Station Illinol's Emerges-cy.Varagement Acercy Division of Nuclear Safety

ATT!%C H, 11\\M1 IF_ I1~'I '

10 CF R '1>u. 6al Regluest for Re1W for Alltarina&(a 11'.egjuM ireinnieinifts, for PressLOire Vessell Inseriioce hispec~IlJoin In ervall Ilii-ii Accoirclaince Vyvflith 110 C1=1111.

R ear u ~s, o ~a ini `il Pagle it of I 1.0 ASME Code Comrponents Ato~ectecl 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.

Descruptiion B-A B1.11 Circumferential Shell Welds B-A B1.21 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 2.0

Iacalb~e Code Edl,t^on and Addenda Tine Third Interval Inservice Inspection (ISI) program is based on the ASME B&PV Code,Section XI, 2001 Edition through 2003 Addenda (Reference 1). T° roughout ;his request the above examination categories are referred to as "the su",) ect examinations" and the ASME B&PV Code,Section XI, is referred to as "the Code."

Tine app'icable Code for the subsequent (i.e., Tourth) IS] interval will be irrp:emented in accordance with the requirements of 10 CFR 50.55x.

`:1

'r=a ica.I~Ie Ctie Rr'Clcirer"P Lnt ASME Section XI Paragraph IWB-2412, "Inspection Program B," requires volumetric and surface examination of essential"y 1 CO% of RPV pressu e retain'.ng welds identified in Table IW3-2500-1 once each '0-year interval. The Braidwood Stat,on 71 rd '0-year iSi tervais are current y scheduled 'to end on Judy 2E, 2013 for nit ` and Cc obey 13, 2013

,or Unt 2.

kt\\, Y'll *?011 I IM11 ILE, IM -if' 110 k; V'11 ; (0. 5 5 a V %1 IL111 E_ F If'~ LE' (C. ~) U Ic,' it Il 3 R -17 lRecluesf for IF',eliief"U`0ir All'Leirnatiive

for R)eaotorr Pressure VesseO lnseiriflce lnspec~ittuoini Ilin

VnMdh A CFR, F age 2 0'u 112 4.0

Reason for Ru us11; 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.

5.0 Profoosed

Use 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

,or 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 :nterval can be extended based on a regligible 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

-;or Category B-A and B-D items, based on a neg:igib?e change In risk, is contained in WCAP-16168-IMP-A, Revision 3, (Reference 4). This methodology was used to develop a pilot plant analysis `or RPVs :n Westinghouse, Combustion Engineering, and Babcock and Wilcox plant designs, and is an extension of the work that was performed as Dart of tine NRC

?ressurized Thermal Shock (PTS) Risk Study (Reference 5). The cr'.t cal parameters for demonstrating t.iat this pilot plant analysis is applicable on a plant specific basis, as identified in WCAP-16168-SIP-A, Revision 3, are s"^own in Table 2. Tab",e 2 lists "he critical parameters investigated in WCAP-16168-NP-A, Rev]sIon 3 and compares the results of the Westinghouse pilot plant to those of Braidwood Units n and 2 for an ISI Interval extension.

Tables 3-'1, 3-2, 4-1, a Id 4-2 provide additional information required by the NRC as noted In App&7cix A of WCA?-161613-NP-A. =revision 3.

By demonstr sting -,-.!a-', eac) pant specific parameter is bounded by the corresponding pilot plant pararr.eter, ~ :-e appllcatio n of the methodology for 3ra:dwood Sl:a"Z:on. Lm,s 1 and 2 is determined to be accep:aaie.

ATTACHMENT 110 GFRI

RELI EF REQUEST 03 -1 e'

,Iequies,l -ioir

for AK(Lernatave Req uurements for Reactor PreSSL rc VQSS~~

ail-useirv~cc l nspectt~oinl ~Nteniaa ~n Accordance ~Aflth 10 CFA Revision 11 Page 3 of 1 Table 1-1

Examination Category, Item Number, Component Description and Component Identification of Applicable Examinations for I3raidWood Unit 1 Exam

Exam Item

component Description

ISI Weld Number Catenory I Numher B-A 131.11 Reactor Vessel Shell-to-Reactor Vessel Shell 1 RV-01-003 Reactor Vessel Shell-to-Reactor Vessel Shell Reactor Vessel Shell-to-Dutchman 1 RV-01-004 1 RV-02-002 131.21 Dutchman-to-Lower Center Disc 1 RV-02-001 131.30 Reactor Vessel-to-Flange 1RV-01-005 B-D B3.90 Outlet Nozzle @ 22 Degrees 1 RV-01-006 Inlet Nozzle @ 67 Degrees 1 RV-01-007 Inlet Nozzle @ 113 Degrees 1 RV-01-008 Outlet Nozzle @ 158 Degrees 1RV-01-009 Nozzle @ 202 Degrees

_Outlet 1 RV-01-010 Inlet Nozzle @ 247 Degrees 1 RV-01-011 Inlet Nozzle a) 293 Degrees

_ 1 RV_-01-012 Outlet Nozzle @ 338 Degrees 1 RV-01-013 B3.100 Outlet Nozzle Inner Radius @ 22 Degrees 1 RV-01-014 Inlet Nozzle Inner Radius @ 67 Degrees 1 RV-01-015 Inlet Nozzle Inner Radius @ 113 Degrees Outlet Nozzle Inner Radius @ 158 Degrees 1 RV-01-016 1 RV-01-017 Outlet Nozzle Inner Radius @ 202 Degrees 1 RV-01-018 Inlet Nozzle Inner Radius @ 247 Degrees 1 RV-01-019 In"et Nozzle inner Radius @ 293 Degrees 11RV-01-020 Outlet Nozzle Inner Radius @ 338 Degrees 1 RV-01-021

A,

'll, f" H Irbil 1'LH: KI1f' 10 C F P-,l 50.15 a Request for ReHef for PUtenrrinlzltove

for 11'GGacll'-olr PlCe-SSL01re VeSSL-~

Anse ce lnspect~on ~ntenfa9 fin, !-\\ccoirdainca YAY th I n CFI P,aga 4~ of I1 2

Table 1-2 ----- - _ -----

Examination Category, Item dumber, Component Description and Component Identification of

/applicable Examinations for Braidwood Unit 6666__

Exam

Exam Item Cate

-gory

Dumber Component Description ISI Weld dumber B-A 131.11 Reactor Vessel Shell-to-Reactor Vessel Shell 2RV-01-003 Reactor Vessel Shell-to-Reactor Vessel Shell 2RV-01-004 2RV-02-002 Reactor Vessel Shell-to-Dutchman 131.21 Dutchman-to-Lower Center Disc 2RV-02-001 131.30 Reactor Vessel-to-Flange 2RV-01-005 B-D B3.90 Outlet Nozzle @ 22 Degrees 2RV-01

-006 Inlet Nozzle @ 67 Degrees 2RV-01-007 Inlet Nozzle @ 113 Degrees 2RV-01-008 Outlet Nozzle

158 Degrees 2RV-01-009 Outlet Nozzle

202 Degrees 2RV-01-010 Inlet Nozzle @ 247 Degrees 2RV-01-011 I n'et Nozzle

293 De rees

2RV-01-0" 2 Outlet Nozzle @ 338 Degrees 2RV-01-013 83.100 Outlet Nozz~le Inner Radius @ 22 Degrees 2RV-01-014 Inlet Nozzle Inner Radius @ 67 Degrees I

2RV-01-015 I Inlet Nozzle Inner Radius (cry 113 Degrees _

_2RV-0`-01,66 Outlet Nozzle Inner Radius @ 158 Degrees 2RV-01-0 17 Outlet Nozzle Inner Radius aD 202 Degrees 2RV-01-0'8 lniet Nozzle Inner Radius @ 247 Degrees 2RV-01-019 Inlet Nozzle Inner Radius @ 293 Decrees 2RV-0'-020 Outlet Nozzle Inner Rad~Ls @ 338 Degrees 2RV-01-021

71

n 7

AlYYACiXl10ilF--li~IT 16 GRI Fr 50.0)1.. a Requiest for F eiieir for AOfieiriniatJve F~Ieqjuiireinnenits for Reactor Pressure Vessel Dnsenfice inspacitioini ~inteirfa~ Iln Accordance With 10 CFRI 50.55a(z)(1 Re~~ie,iiou~

'P'age 6 of I Table Critical Parameters for the Application of Bounding Analysis for Braidwood Units 'I and 2 Parameter Pilot Plant Basis Plant-Specific Basis Additional Evaluation Required Dominant Pressurized NRC PTS Risk Study

PTS Generalization No Thermal Shock (PTS)

(Reference 7) Study (Reference 6)

Transients in the NRC PTS Risk Study are Applicable Through-Wall Cracking 1.76E-08 Events Unit 1: 2.09E-16 Events No Frequency (TWCF) per year per year (calculated per (Reference 4)

Reference 4) f Unit 2: 2.04E-15 Events per year (calculated per Reference 4)

Frequency and Severity of 7 heatup / cooldown

Bounded by 7 No Design Basis Transients cycles per year heatup / cooldown (Reference 4) cycles per year C;adding Layers

Single Layer Single Layer No (Singie/Multiple)

(Reference 4)

ATTACHMENT 10 CFR1 60e56a RE lEF REQUEST 0381-1 e' R,egluiest ifoir Relief for Alternative Requirements for Reactor P ressuiire Nfe'sel iuwseiMce lInspection lnteirval nn Accordance With 10 CIFI,71 Revision I Page 6 of 12 Tables 3-1 and 3-2 provide a surnmary 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 smell 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 1/10"' or 1 inch of the reactor vessel thickness; therefore, they are inherently acceptable per the requirements of the Alternate PTS Ru:e,

~

10 CFR 50.61 a (Reference 7).

Proposed :nspection The third inservice inspection currently is sc:ieduied for 2018. This inspection schedu:e for balance of

will be performed in 2027 plus or minus one refue'ing outage. The p:ant life:

proposed inspection date for Braidwood Urit 1 is consistent with,he latest I 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 1 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 1/10`x' 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

Number of

......... --..--.-- ------------------------------------- I....._..

Extent, TWE maximum forging flaws number of TWEMIN

TWEMAx

(Axial/Circ.)

forging flaws 0.075

0.375 76

1 (0/1) 0.125 0.375 30 1 (011) 0.375 8

0(0/0) 0.175 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 1U CFR U. 5dt RELIEF REQUEST MR-17 K cjuest for

for Alternative Requirements for Reactor Pressure Vessel Inservicc Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)

Revision 1 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; TRC,S [°F]:

N/A Twall [inches]:

8.025 No.

Region and Component Description Material Heat No.

Cum

[~0]

NP )

(;ti'/o]

R.G.0) 99 Pos.

1.

° CF(1)

[F]

RT~oT~~~(1)

[OF]

Fluence [10 i-r cM, E>1 0 P.eV; ( 1 '

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

,r.

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 Region No. (From Above)

RTI,,Ax-xx [°R]

Fluence

[10 1'9 n/cmZ, E > 1.0 MeV]

FF (Fluence Factor)

AT;_ [°F]

TWCF; 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(aFOTW C F95-FO +acwTWC F95-CW):

2.09E-16 (1) Reference 8 (2) Reference 9

ATTACH I( TENT 10 CFIR bO.66a u ELIEF REQUEST 13R-17 acst for Relief for Alternative Requirements for Reactor Pressure.Vessel Inservice Inspection Interval In Accordance kMth 10 CFR 50.55a(z)(1)

Revision 1 Gage 9 of 12 Details of TWCF Calculation for Braidwood Unit 2 at 57 Effective Full-Power Years (EFPY)

Inputs Reactor Coolant System Temperature, TRcs [°F]:

N/A T,;

[inches]:

8.625 No.

Region and Component Description Material Heat No.

Cum

[wt%]

[v.!%]

Ni(1)

R.G.(1) 1.99 POs.

CF(1)

[°F]

RT,;pTtuy(1) [~F]

Fluence [10'V n;cm'.

E>1.0 MeV] (1) 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 AT30:

Regulatory Guide 1.99. Revision 2(2)

Controlling Material Region No. (From Above)

RT,,.

.^

[~R]

Fluence

[10 19

2 E > 1.0 MeV]

FF (Fluence Factor)

AT,,3

,:,3 [°F]

TWCF,=

Limiting Forging - FO 1

516.43 1.11 1.029 26.76 8.18E Limiting Circumferential Weld - CW 5

552.69 3.03 1.293 53.02 0.00E+00 TWCF.;-ToT _L(aFoTWCF,5_F0+ a,,..;TWCF

y}:

2 04E-15 (1) Reference 8 (2) Reference 9

~~7i"l~'~CHMENT 110 CFR" 1.5,0.di5er IRELI E FF IREQUE.

3-`Al-11'x' Requiesk for Relief for

Reglruiiirennen-t ~ ifor R`e,-Eici'tor I'

.~

~"resa!iir~:

e~S VSell Inservice Inslpecition Interved h-,i ~~\\ccoirdance ~~~9ii~ah~ '110 I~e~~isioin~ 11 Page 110 of 12 6.0

our '(Jon of, Pr otiose l.._

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

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-U1-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 ASVE 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. I31R-08 and 13R-09 cor the Tnlyd 10-Year Inservice Inspection Program Interval (TAC Nos. MF3321 and MF3322)," dated December 10, 2014 (ADAMS Accession Number ll1L14321A864)

8. "Watts Bar Nuclear Plant, Unit 1-Request for Alternative ' 3-ISI-01 to Extend the Second Reactor Vessel Weld "nservice Inspection Interval (TAC No. AF2958)," dated November 28, 2014 (ADAVIS Accession Number ML14314A987)

9. "Sequoyah Nuclear Plant, "_Xits 1 and 2 Requests for Alternatives 13-ISI-1 ard 13-:SI-2 L 'extend the Reactor Vessel Weld Inservice °rspecon Interval (TAC Nos. N =2900 and

.V,F2901)," dated August ", 2014 (ADAMS Accession number JVL14188B920)

10. `Catawba Nuclear Station Jnits 1 a;rd 2: Proposed Relief Request 13-CN-0,03, Request or X ernative to the Requ]rerr.°ent of lWB-2500, Table IWB-2500-1, Category B-A and Category 3-D for Reactor Pressure Vessel Weds (TAC Nos. V,,F1922 ar'd VIF1923),"

,'a-'-ed Varcii 23, 2.014 (ADAMS Accession Number. L14079A546)

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST 13R-17 IRegUest fOV I~Zelief for Alternative Requirements for Reactor Pressure Vessel lInseirvice (inspection Interval In Accordance With 10 CFR 50.55a(z)(1)

Revision 1 Pagel 1 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 (ADAIVIS Accession Number IVIL14198A331)

13. "Sorry 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 I nspection 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 Ore Outage) (TAC No. ME3010)," dated July 12, 2010 (ADAMS Accession Number MI-101750402)

18. "Palo Verde Nucear Gererating Station, Units 1, 2, and 3 Relief Request No. RR-40, :Reactor Vessel Weld Examination Interval Extension (TAC Nos. ME1634, ME1635, and W.E1636)," dated February 22, 2010 (ADAMS Accession Number ML1 CO290415)

19. "Safety Evaluation of Relief Requests to Extend the Inservice Irspection'n-ervai for Reactor Vesse Examirations for Sa'em Nuclear Gererating Station, Unit Nos. 1 and 2 (TAC Nos. ME1478, ME1479, ME1480 and M-1481)," dated February 22, 2010 (ADAMS Accession Number ML10049155C)

ATTACHMENT 10 CFR 50.55a RELIEF REQUEST I3R-17 U_,equest for Relief for Alternative Requirements for Reactor (Pressure Vessell Inservice Inspection Interval In Accordance With 10 CFR 50.55a(z)(1)

Revision 1 Page 12 of 12 8.0 G.~~~i_~rainc2s 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)

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 w th 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 ' and 2 Reactor Vessel Integrity Evaluation to Support Licerse Renewal Time-Limited Aging Analysis," December 2012

9.

NRC regulatory Guide 1.99, Revision 2, "radiation Ersbrittlement of Reactor Vessel iVaterials," May 1988

,, 0. OG-06-356, "Man for Plant-Specific Implementation of Extended Inservice Inspection interval per WCAP-16138-NP, Revision 1, `Risk Informed Extension of tine reactor Vessel in-Service Inspection Interval,' 1 UHP 5097-99, 'ask 2059," October 31, 2006