NL-12-1446, Proposed Inservice Inspection Alternatives VEGP-ISI-AL T-05 and VEGP-ISI-AL T -06 for Implementation of Extended Reactor Vessel Inservice Inspection Intervals

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Proposed Inservice Inspection Alternatives VEGP-ISI-AL T-05 and VEGP-ISI-AL T -06 for Implementation of Extended Reactor Vessel Inservice Inspection Intervals
ML12243A248
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 08/29/2012
From: Ajluni M
Southern Co, Southern Nuclear Operating Co
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
NL-12-1446
Download: ML12243A248 (20)
v • d • e


Text

Mark J. Ajluni, P.E. Southern Nuclear Nuc lear Licensing Director Operating Company, Inc.

40 Invern ess Center Parkway Post Office Box 1295 Binningham. Alabama 35201 Tel 205.992.7673 Fax 205.992.7885 August 29, 2012 SOUTHERN'\'

COMPANY Docket Nos.: 50-424 NL-12-1446 50425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 & 2 Proposed Inservice Inspection Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT -06 for Implementation of Extended Reactor Vessel Inservice Inspection Intervals Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(a)(3)(i) , Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (lSI) Alternatives VEGP-ISI-AL T-05, Version 1 and VEGP-ISI-ALT 06, Version 1. These alternatives propose one-time extensions of reactor vessel examinations from a ten-year examination period to a twenty-year examination period for their respective unit. Specifically, these alternatives request one-time extensions to selected ASME Examination Category 8-A reactor vessel welds and selected ASME Examination Category 8-D reactor vessel nozzle welds and reactor vessel inside radius sections.

The approval of VEGP-ISI-ALT-05 will move examinations currently scheduled for Unit 1 Reload 19 (1 R19, Fall 2015) to either 1R25 (Fall 2024) or 1R26 (Spring 2026). The approval of VEGP-ISI-ALT-06 will move examinations currently scheduled for Unit 2 Reload (2R18, Spring 2016) to either 2R24 (Spring 2025) or 2R25 (Fall 2026). Approval of these alternatives is requested by September 1, 2013.

This letter contains no NRC commitments. If you have any questions, please contact me at (205) 992-7673.

Respectfully submitted, M. J. Ajluni Nuclear Licensing Director MJA/RMJ/lac

U. Nuclear Regulatory Commission NL-1 446 Page 2

1. Proposed Alternative VEGP-ISI-ALT-05, 1.0, in With 10 50.55a(a)(3)(i)
2. Proposed Alternative VEGP-ISI-ALT-06, Version 1.0, in Accordance With 10 CFR 50.55a(a)(3)(i) cc:

Kuczynski, President & CEO Executive Vice President & Chief Nuclear Officer Mr. T. E. Vice President - Vogtle Mr. L. lvey, Vice President - Regulatory Affairs Mr. J. Adams, Vice President Fleet Operations CVC7000 Mr. V. M. Regional Administrator Mr. R. E. Martin, NRR Senior Project Manager - Vogtle Mr. L. M. Cain, Senior Resident Inspector - Vogtle

Vogtle Electric Generating Plant - Unit 1 Proposed Alternative for the Third lSI Interval Enclosure 1 Proposed Alternative VEGP-ISI-AlT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

Enclosure 1 Proposed Alternative VEGP-ISI-AL T-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

Plant Site-Unit: Vogtle Electric Generating Plant (VEGP) - Unit 1.

Interval Dates: 3rd Inservice Inspection (lSI) Interval- May 31,2007 through May 30,2017.

Requested Date for Approval: Approval is requested by September 1, 2013.

The affected components are Examination Category 8-A, Items 81.11, 81.12, ASME Code 81.21,81.22,81.30, and 81.40 reactor vessel (RV) welds, and Examination Components Affected: Category 8-0, Items 83.90 and 83.100 RV nozzle welds and nozzle inside radius section. The specific components are provided in Table 4.

Applicable Code The applicable code edition and addenda (for the 3rd lSI interval) is ASME Section Edition and XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Addenda: Edition with the 2003 addenda (Reference 1).

Applicable Code Table IW8-2500-1 requires volumetric examination of the affected RV components Requirements: once each ten-year interval.

The Westinghouse pilot plant RV analysis defined in WCAP-16168-NP-A, Revision 3 (Reference 2) utilizes probabilistic fracture mechanics and risk analysis methods to justify extending the lSI interval for reactor vessel welds (Examination Category 8-A) and nozzle-to-vessel welds and nozzle inside radius section (Examination Category 8-0) from 10 years to 20 years.

Reason for Request:

An analysis has been performed showing that VEGP - Unit 1, which is a Westinghouse 4-Loop plant, is bounded by the pilot plant parameters defined in Reference 2. Therefore, Southern Nuclear Operating Company (SNC) is requesting approval of this alternative to allow the use of the lSI interval extension for the affected VEGP - Unit 1 components.

SNC is requesting a one-time extension of the lSI interval from 10 years to 20 years for VEGP - Unit 1 Examination Category 8-A reactor vessel welds and Examination Category 8-0 nozzle-to-vessel welds and nozzle inside radius section.

Proposed Specifically, this proposed alternative would permit the deferral of the ASME Code Alternative:

required Examination Category 8-A and 8-0 volumetric examinations currently scheduled for the Fall of 2015 (3rd period of 3rd interval) until 2026, plus or minus one refueling cycle (3rd period of 4th interval). The required examinations would subsequently be performed using the Section XI Code in effect for the 4th interval.

E1-1

Enclosure 1 Proposed VEGP-ISI-ALT-OS Version 1.0, in Accordance with 10 CFR SO.S5a{a){3){i)

Basis for Use: The methodology to demonstrate the acceptability of extending the intervals Category B-A and components is contained in WCAP 16168-NP-A, 3 (Reference was used to a pilot plant risk for Westinghouse rt!..), Engineering Babcock and Wilcox (B&W) RV designs and is an of the work that was performed as of the NRC Pressurized Thermal Shock (PTS) Risk (Reference Reference 2 used the through wall cracking frequency (TWCF) as a measure of the risk of RV failure, and it was demonstrated that inspection interval for the affected components can be extended from 10 to 20 years while the change in risk guidelines found in Regulatory Guide 1.174 (Reference 4).

Reference 2 was approved by the NRC in a July 2011 safety evaluation (ML111 3.4 of the evaluation provides the for a utility to an in with 10 CFR 50.55a(a)(3)(i) to use the Reference 2 for a plant specific evaluation. requirements are below:

1. Licensees demonstrate that the embrittlement of their RV is within envelope in the supporting A plant specific analysis was performed that demonstrated that Unit 1 RV are bounded by corresponding pilot plant parameters. The critical are identified in Table 1. Table 3 provides detailed information relative to the calculation of
2. whether limiting design during prior operation are than the frequency PWROG 2) fatigue analysiS.

As shown in 1, the frequency of - Unit 1 limiting design are bounded by the frequency OnTlTlCln in the PWROG fatigue analysis.

3. must report the results of prior lSI of RV welds and the proposed schedule for the next 20 year interval.

of the previous RV inspections for VEGP - Unit 1 are provided in information examinations have been nOI"fr.r'm"'n on the VEGP - Unit 1

4. for an alternative, I'Clr~"'.c,O shall identify the in which inspections will be no,r1't'\,'m",ri The Unit 1 RV examinations currently scheduled for 2015 will be deferred until plus or minus one refueling cycle. The dates provided be within plus or minus one refueling cycle of date identified in PWROG letter OG-1 0 238, Jul 12, 2010 Reference 5 .

E1-2

Enclosure 1 Proposed Alternative VEGP-ISI-AlT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

- Unit 1 is bounded by pilot plant application because the total TWCF for

- Unit 1 was calculated as 7.66E-1 This value is less than the value of which was calculated for the Westinghouse pilot plant in 2; therefore, the use of this proposed alternative will provide an acceptable level quality and safety. Therefore, it is requested that the NRC authorize this proposed alternative in accordance with 10 50.55a(a)(3)O).

Duration of Proposed The 3rd and 4th Intervals.

Alternative:

Precedents:

1. Donald Cook Nuclear Plant, Unit 2 Evaluation of Relief Request (ISIR-29) to the Third i0-Year lnservice Inspection Interval for Reactor Weld Examination (TAC MD9934) Donald C. Cook Nuclear Power Plant, Unit 2 - NRC Safety Report dated June 8,2009 (ML091260163).

Safety Evaluation for Relief Requests ISI-090 & 021 Reactor Vessel Weld Examination Extension Calvert Cliffs Nuclear Power Plant, Unit 2 (TAC Nos.

MD9773 and MD9774). Calvert Cliffs Nuclear Power Plant Unit No.2 NRC Safety Evaluation Report April 8, 2009 (ML090920077).

Palisades Plant - Evaluation of Request to Extend the Third 1 Inservice Inspection Interval Reactor Weld Examinations (TAC NO. MD9265)

Plant NRC Safety Report dated February 11, 2009 (ML090120896).

4. Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No.1 Reactor Weld Examination Extension (TAC NO. MD9962) - R.E. Ginna Plant NRC Report dated July 31, 2009 (ML092080229).

Kalyanam, N., NRC, to Vice President, Entergy Operations, 'Waterford Steam Electric Station, Unit 3 - Withdrawal of an Amendment Request (TAG NO.

MD9669)," dated June 1 2009 (ML0916001 and ML091600158).

6. Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) Relief Request for Extension of the Reactor Vessel Inservice Inspection date to year 2020 (Plus or Minus Outage) (TAC NO. ME3010). NRC Safety Evaluation Report dated July 1 0 (ML101750402).

References:

1. American Society of Mechanical Engineers. ASME Boiler and Pressure Code,Section XI, 2001 Edition with the 2003 Addenda.
2. WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of Reactor Vessel In-Service Ins Interval," October 2011. roved under NRC

Enclosure 1 Proposed Alternative VEGP-ISI-AlT-05 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

ML111600303}.

3. NUREG-1 "Recommended Screening Limits for Pressurized Thermal Shock," March 2010.
4. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing " November 2002.

OG-10-238, "Revision the Plan for Plant Specific Implementation of Extended Inservice Inspection Interval perWCAP-i6168-NP, Revision 1, "Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval.' PA MSC-0120," July 1 O.

NRC Letter Report, of Plant-Specific Pressurized Thermal Shock (PTS) Risk to Additional Plants," December 14, 2004.

Nuclear Regulatory Commission, of Regulations, 10 CFR Part 50.6ia, "Alternate Fracture Toughness Requirements for Protection against Thermal Shock " Washington D. Federal Register, Volume 75, No.1, January 4, 2010 and No. 22 with corrections to part (g)

February 3, 2010, March 2010, November 26, 2010.

8. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
9. WCAP-1 Revision "Vogtle Unit 1, Heatup and Cooldown Limit Curves for Normal Operation," July 2009.

Awaiting NRC approval.

-4

Enclosure 1 Proposed Alternative VEGP-ISI-AlT-OS Version 1 in Accordance with 10 SO.SSa(a)(3)(i)

Table 1 Critical Parameters for the Application of the Bounding Analysis as Applied to VEGP - Unit 1 Additional VEGP- Unit 1 Evaluation Parameter Pilot Plant Basis Basis Required?

Dominant Pressurized Thermal NRC Risk Study PTS Generalization No Shock (PTS) Transients in the NRC (Reference 3) Study (Reference 6)

PTS Risk Study are applicable Through-Wall Cracking Frequency 1.76E-08 Events per 7.66E-14 Events per No (TWCF) year (Reference year (Calculated using Reference 2)

Frequency and Severity of Design 7 heatup/cooldown Bounded by 7 No I Basis Transients

  • cycles per year heatup/cooldown

! (Reference 2) cycles per year (1)

Cladding Layers (SingleIMultiple) Single Layer No (Refel 2)

Note:

(1) Per VEGP Application for License Renewal, 60 years of operation, the projected number of design transients is below the number specified in the 40-year bases. a result, VEGP Unit 1 is conservatively bounded by 7 heatup!cooldown E1

Enclosure 1 Proposed Alternative VEGP-ISI-ALT-OS Version 1.0, in Accordance with 10 CFR SO.SSa(a)(3)(i)

Table 2 Additional Information Pertaining to Reactor Vessel Inspections for VEGP - Unit 1

. Inspection The latest lSI was conducted in accordance with the ASME Code,Section XI 1989 I methodology: Edition. Examinations of Category B-A and B-D welds were performed to Section XI Appendix VIII, 1995 Edition with the 1996 Addenda, as modified 10 CFR 50.55a(b)(2)(xiv, xv and xvi). Future inservice inspections will performed to ASME Section XI Appendix VIII requirements.

Number of past Two i0-Year inservice inslpec:tlorls have been performed.

inspections:

Number of There were two indications identified in the beltline region during the most recent indications found: inservice inspection. These indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Only one indication is within the inner 1/1Oth or 1 inch of the reactor vessel thickness. This indication is located the axial weld fusion line but is conservatively considered to be in intermediate shell B8805-3. This indication is per the requirements of the Alternate Rule, 10 CFR 50.61a HA'IArI::onr'A 7), since the flaw is less than the allowable number of flaws for each flaw size increment. A disposition of flaw against the limits of Alternate Rule is shown in the table below.

Through-Wall Extent, TWE (in) Number of maximum plate flaws number of (Axial/Circ, )

TWEMIN TWEMAX plate flaws 0.075 0.375 73 1/0) 0.125 0.375 29 1 (i/O) 0.175 0.375 8 1 (1/0) 0.225 3 1 (1/0) 0.275 0.375 1 1 (1/0)

This indication is 1.6 inches in 0.32 inches in through-wall extent dimension),

and is embedded with an'S' dimension of 0.55 inches as measured from the cladding-to base-metal interface.

This indication was observed in the second lSI interval inspection only; it was not recorded during the first lSI interval examinations. The second lSI interval examination was an Appendix VIII (PDI) examination with a much higher sensitivity than the interval Section XI/Regulatory 1.150 examination.

Proposed The third inservice is scheduled for 2015. This will be performed in schedule for balance or minus one refueling The DroiDo~:;ea InSIJeCtiOn date is consistent with of life: . the latest revised implementation OG-10-238 (Reference plus or minus one

! refueling

Enclosure 1 Proposed Alternative VEGP-ISI-AlT-OS Version 1.0, in Accordance with 10 CFR SO.SSa(a)(3)(i)

Table 3 Details of the Throu h-Wall Crackin Reactor Coolant System Temperature. T dOF]: T wall [inches]: 8.781 Cu(1) Ni(l) (1) Fluence

! No. and Component Material RTNDT(u) 19 2

[10 Neutron/cm ,

Description Heat No. [wt%] [wt%] e'F]

E> 1.0 MeV]

1 Intermediate Sheil Plate C-0613-1 0.083 0.597 0 3.53 2 Intermediate Shell Plate 88805-2 C-0613-2 0.083 0.610 20 3.53 3 Intermediate Plate 88805-3 C-0623-1 0.062 0.598 2.1 47.0 30 4 Lower Shell Plate 88606-1 0.593 1.1 32.8 20 3.53

! 5 Lower Shell Plate 88606-2 0.60 1.1 35.2 20 3.53 6 Lower Shell Plate 88606-3 41.9 10 3.53 Weld 101-124A 23.3 -80 1.67 Weld 101-1248 23.3 -80 3.17 3.17

-80 3.17 Shell Long. Weld 101-1428 -80 1.67 12 Lower Shell Long. Weld 101 83653 0.042 0.102 2.1 23.3 -80 3.17 Inter. To Lower Shell Weld 13 83653 0.042 0.102 2.1 23.3 -80 3.53 101-171 Outputs Methodology Used to Calculate Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Fluence FF Material RT MAX.XX 2

[OR] (1 Neutron/cm , (Fluence [oF]

No.

E> 1.0 MeV]

Above)

Limiting Axial Weld AW 3 550.96 3.17 Limiting Plate - PL 3 552.12 3.53 Circumferential Weld CW 3 552.12 3.53 TWCFs5-TOTAL(aAWTWCF 95*AW + apLTWC Note:

(1) Reference 9

-7

Enclosure 1 Proposed Alternative VEGP-ISI-AL T-OS Version 1.0, in Accordance with 10 CFR SO.S5a(a)(3)(i)

Table 4 List of Affected Components for VEGP - Unit 1 ASME ASME ITEM COMPONENT ID DESCRIPTION CATEGORY NUMBER B-A B1.11 11201-V6-001-W04 UPPER SHELL TO INTERMEDIATE SHELL WELD B-A B1 .11 11201-V6-001-W05 INTERMEDIATE SHELL TO LOWER SHELL WELD B-A B1.11 11201-V6-001-W06 LOWER SHELL TO BOTTOM HEAD TORUS WELD B-A B1.12 11201-V6-001-W12 UPPER SHELL LONGITUDINAL WELD AT 42 DEGREES B-A B1.12 11201-V6-001-W13 UPPER SHELL LONGITUDINAL WELD AT 162 DEGREES B-A B1.12 11201-V6-001-W14 UPPER SHELL LONGITUDINAL WELD AT 282 DEGREES B-A B1.12 11201-V6-001-W15 INTERMEDIATE SHELL LONGITUDINAL WELD AT 0 DEGREES B-A B1.12 11201-V6-001-W16 INTERMEDIATE SHELL LONGITUDINAL WELD AT 120 DEGREE B-A B1.12 11201-V6-001-W17 INTERMEDIATE SHELL LONGITUDINAL WELD AT 240 DEGREE B-A B1.12 11201-V6-001-W18 LOWER SHELL LONGITUDINAL WELD AT 60 DEGREES B-A B1.12 11201-V6-001-W19 LOWER SHELL LONGITUDINAL WELD AT 180 DEGREES B-A B1.12 11201-V6-001-W20 LOWER SHELL LONGITUDINAL WELD AT 300 DEGREES B-A B1.21 11201-V6-001-W01 CLOSURE HEAD DOME TO TORUS WELD B-A B1.21 11201-V6-001-W07 BOTTOM HEAD TORUS TO BOTTOM HEAD DOME WELD B-A B1.22 11201 -V6-001-W08 CLOSURE HEAD TORUS MERIDIONAL WELD AT 45 DEGREES B-A B1 .22 11201-V6-001-W09 CLOSURE HEAD TORUS MERIDIONAL WELD AT 135 DEGREES B-A B1.22 11201-V6-001-W10 CLOSURE HEAD TORUS MERIDIONAL WELD AT 225 DEGREES B-A B1.22 11201-V6-001-W11 CLOSURE HEAD TORUS MERIDIONAL WELD AT 315 DEGREES B-A B1.22 11201-V6-001-W21 BOTTOM HEAD TORUS MERIDIONAL WELD AT 0 DEGREES B-A B1.22 11201-V6-001-W22 BOTTOM HEAD TORUS MERIDIONAL WELD AT 90 DEGREES B-A B1.22 11201-V6-001-W23 BOTTOM HEAD TORUS MERIDIONAL WELD AT 180 DEGREES B-A B1.22 11201-V6-001-W24 BOTTOM HEAD TORUS MERIDIONAL WELD AT 270 DEGREES B-A B1.30 11201-V6-001-W03 VESSEL SHELL TO FLANGE WELD B-A B1.40 11201-V6-001-W02 CLOSURE HEAD TORUS TO FLANGE WELD B-D B3.90 11201-V6-001-W25 VESSEL TO OUTLET NOZZLE N1 WELD AT 22 DEGREES B-D B3.90 11201-V6-001-W26 VESSEL TO INLET NOZZLE N2 WELD AT 67 DEGREES B-D B3.90 11201-V6-001-W27 VESSEL TO INLET NOZZLE N3 WELD AT 113 DEGREES B-D B3.90 11201-V6-001-W28 VESSEL TO OUTLET NOULE N4 WELD AT 158 DEGREES B-D B3.90 11201-V6-001-W29 VESSEL TO OUTLET NOZZLE N5 WELD AT 202 DEGREES B-D B3.90 11201-V6-001-W30 VESSEL TO INLET NOZZLE N6 WELD AT 247 DEGREES B-D B3.90 11201-V6-001-W31 VESSEL TO INLET NOZZLE N7 WELD AT 293 DEGREES B-D B3.90 11201-V6-001-W32 VESSEL TO OUTLET NOZZLE N8 WELD AT 338 DEGREES B-D B3.100 11201-V6-001-IR01 OUTLET NOZZLE N1 INNER RADIUS AT 22 DEGREES B-D B3.100 11201-V6-001-IR02 INLET NOZZLE N2 INNER RADIUS AT 67 DEGREES B-D B3.100 11201-V6-001-IR03 INLET NOZZLE N31NNER RADIUS AT 113 DEGREES B-D B3.100 11201-V6-001-IR04 OUTLET NOZZLE N41NNER RADIUS AT 158 DEGREES B-D B3.100 11201-V6-001-IR05 OUTLET NOZZLE N5 INNER RADIUS AT 202 DEGREES B-D B3.100 11201-V6-001-IR06 INLET NOZZLE N6 INNER RADIUS AT 247 DEGREES B-D B3.100 11201-V6-001-IR07 INLET NOZZLE N7 INNER RADIUS AT 293 DEGREES B-D B3.100 11201-V6-001 -IR08 OUTLET NOZZLE N81NNER RADIUS AT 338 DEGREES E1-8

Vogtle Electric Generating Plant - Unit 2 Proposed Alternative for the Third 151 Interval Enclosure 2 Proposed Alternative VEGP-ISI-AlT-06 Version 1 in Accordance with 10 CFR 50.55a(a)(3)(I)

Enclosure 2 Proposed Alternative VEGP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

Plant Site-Unit: Vogtle Electric Generating Plant (VEGP) - Unit 2.

Interval Dates: 3rd Inservice Inspection (lSI) Interval - May 31,2007 through May 30, 2017.

Requested Date for Approval : Approval is requested by September 1,2013.

The affected components are Examination Category 8-A, Items 81.11, 81.12, ASME Code 81.21, 81.22, 81.30, and 81.40 reactor vessel (RV) welds, and Examination Components Affected: Category 8-0, Items 83.90 and 83.100 RV nozzle welds and nozzle inside radius section. The specific components are provided in Table 4.

Applicable Code The applicable code edition and addenda (for the 3rd lSI interval) is ASME Section Edition and XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Addenda: Edition with the 2003 addenda (Reference 1).

Applicable Code Table IW8-2500-1 requires volumetric examination of the affected RV components Requirements: once each ten-year interval.

The Westinghouse pilot plant RV analysis defined in WCAP-16168-NP-A, Revision 3 (Reference 2) utilizes probabilistic fracture mechanics and risk analysis methods to justify extending the lSI interval for reactor vessel welds (Examination Category 8-A) and nozzle-to-vessel welds and nozzle inside radius section (Examination Category 8-0) from 10 years to 20 years.

Reason for Request:

An analysis has been performed showing that VEGP - Unit 2, which is a Westinghouse 4-Loop plant, is bounded by the pilot plant parameters defined in Reference 2. Therefore, Southern Nuclear Operating Company (SNC) is requesting approval of this alternative to allow the use of the lSI interval extension for the affected VEGP - Unit 2 components.

SNC is requesting a one-time extension of the lSI interval from 10 years to 20 years for VEGP - Unit 2 Examination Category 8-A reactor vessel welds and Examination Category 8-0 nozzle-to-vessel welds and nozzle inside radius section.

Proposed Specifically, this proposed alternative would permit the deferral of the ASME Code Alternative:

required Examination Category 8-A and 8-D volumetric examinations currently scheduled for the Spring of 2016 (3rd period of 3rd interval) until 2026, plus or minus one refueling cycle (3rd period of 4th interval). The required examinations would subsequently be performed using the Section XI Code in effect for the 4th interval.

E2-1

Enclosure 2 Proposed Alternative VEGP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

Basis for Use: The methodology used to demonstrate the acceptability of extending the inspection intervals for Examination Category B-A and B-O components is contained in WCAP 16168-I\lP-A, Revision 3 (Reference 2). This methodology was used to develop a pilot plant risk analysis for Westinghouse ('!{), Combustion Engineering (CE), and Babcock and Wilcox (B&W) RV designs and is an extension of the work that was performed as part of the NRC Pressurized Thermal Shock (PTS) Risk Re-Evaluation (Reference 3). Reference 2 used the estimated through wall cracking frequency (TWCF) as a measure of the risk of RV failure, and it was demonstrated that the inspection interval for the affected components can be extended from 10 years to 20 years while meeting the change in risk guidelines found in Regulatory Guide 1.174 (Reference 4).

Reference 2 was approved by the NRC in a July 26,2011 safety evaluation (ML111600303). Section 3.4 of the safety evaluation provides the requirements for a utility to submit an alternative in accordance with 10 CFR SO.S5a(a)(3)(i) to use Reference 2 for a plant specific evaluation. These requirements are addressed below:

1. Licensees must demonstrate that the embrittlement of their RV is within the envelope used in the supporting analyses.

A plant specific analysis was performed that demonstrated that VEGP - Unit 2 RV parameters are bounded by corresponding pilot plant parameters. The critical parameters are identified in Table 1. Table 3 provides detailed information relative to the calculation of the TWCF.

2. Licensees must report whether the frequency of the limiting design basis transients during prior operation are less than the frequency identified in the PWROG (Reference 2) fatigue analysis.

As shown in Table 1, the frequency of the VEGP - Unit 2 limiting design basis transients are bounded by the frequency identified in the PWROG (Reference 2) fatigue analysis.

3. Licensees must report the results of prior lSI of RV welds and the proposed schedule for the next 20 year lSI interval.

The results of the previous RV inspections for VEGP - Unit 2 are provided in Table 2. This information confirms that satisfactory examinations have been performed on the VEG P - Unit 2 RV.

4. In the request for an alternative, each licensee shall identify the years in which the future inspections will be performed.

The VEGP - Unit 2 RV examinations currently scheduled for 2016 will be deferred until 2026, plus or minus one refueling cycle. The dates provided must be within plus or minus one refueling cycle of the date identified in PWROG letter OG-1 0 238, dated July 12, 2010 (Reference 5).

E2-2

Enclosure 2 Proposed Alternative VEGP-ISI-AL T-06 Version 1.0, in Accordance with 10 CFR SO.SSa(a)(3)(i)

VEGP - Unit 2 is bounded by the pilot plant application because the total TWCF for VEGP - Unit 2 was calculated as 1.22E-13. This value is less than the value of 1.76E-08 which was calculated for the Westinghouse pilot plant in Reference 2; therefore, the use of this proposed alternative will provide an acceptable level of quality and safety. Therefore, it is requested that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(a)(3)(i).

Duration of Proposed The 3rd and 4th Intervals.

Alternative:

Precedents:

1. Donald C. Cook Nuclear Plant, Unit 2 - Evaluation of Relief Request (ISIR-29) to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examination (TAC MD9934) Donald C. Cook Nuclear Power Plant, Unit 2 - NRC Safety Evaluation Report dated June 8, 2009 (ML091260163) .
2. Safety Evaluation for Relief Requests ISI-090 & 021 Reactor Vessel Weld Examination Extension - Calvert Cliffs Nuclear Power Plant, Unit 2 (TAC Nos.

MD9773 and MD9774). Calvert Cliffs Nuclear Power Plant Unit NO.2 - NRC Safety Evaluation Report dated April 8, 2009 (ML090920077).

3. Palisades Plant - Evaluation of Relief Request to Extend the Third 10-Year Inservice Inspection Interval For Reactor Weld Examinations (TAC 1\10. MD9265)

Palisades Plant - NRC Safety Evaluation Report dated February 11, 2009 (ML090120896).

4. R.E. Ginna Nuclear Power Plant: Safety Evaluation for Relief Request No. 18, Reactor Vessel Weld Examination Extension (TAC NO. MD9962) - R.E. Ginna Plant - NRC Safety Evaluation Report dated July 31,2009 (ML092080229).
5. Kalyanam, N., I\IRC, to Vice President, Entergy Operations, 'Waterford Steam Electric Station, Unit 3 - Withdrawal of an Amendment Request (TAC NO.

MD9669)," dated June 12, 2009 (ML091600132 and ML091600158).

6. Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of the Reactor Vessellnservice Inspection date to the year 2020 (Plus or Minus One Outage) (TAC NO. ME3010), NRC Safety Evaluation Report dated July 12, 2010 (ML101750402).

References:

1. American Society of Mechanical Engineers. ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition with the 2003 Addenda.
2. WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," October 2011. (Approved under NRC SER E2-3

Enclosure 2 Proposed Alternative VEGP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

ML111600303)

3. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock," March, 2007.
4. 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," November 2002.
5. 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.
6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004.
7. Nuclear Regulatory Commission, Code of Federal Regulations, 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. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
9. WCAP-17353-NP, Revision 0, "Vogtle Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," August 2011 .

Status: Awaiting NRC approval.

E2-4

Enclosure 2 Proposed Alternative VEGP-ISI-AL T-06 Version 1.0, in Accordance with 10 CFR SO.SSa(a}(3}(i}

Table 1 Critical Parameters for the Application of the Bounding Analysis as Applied to VEGP - Unit 2 Additional VEGP - Unit 2 Evaluation Parameter Pilot Plant Basis Basis Required?

Dominant Pressurized Thermal I\IRC PTS Risk Study PTS Generalization No Shock (PTS) Transients in the NRC (Reference 3) Study (Reference 6)

PTS Risk Study are applicable Through-Wall Cracking Frequency 1.76E-08 Events per 1.22E-13 Events per No (TWCF) year (Reference 2) year (Calculated using Reference 2)

Frequency and Severity of Design 7 heatup/cooldown Bounded by 7 No Basis Transients cycles per year heatup/cooldown (Reference 2) cycles per year (1)

Cladding Layers (Single/Multiple) Single Layer Single Layer No (Reference 2)

Note:

(1) Per the VEGP Application for License Renewal, after 60 years of operation, the projected number of design basis transients is below the number specified in the 40-year design bases. As a result, VEGP - Unit 2 is conservatively bounded by 7 heatup/cooldown cycles per year.

E2-5

Enclosure 2 Proposed Alternative VEGP-ISI-AlT-06 Version 1.0, in Accordance with 10 CFR SO.SSa(a)(3)(i)

Table 2 Additional Information Pertaining to Reactor Vessel Inspections for VEGP - Unit 2 Inspection The latest lSI was conducted in accordance with the ASME Code,Section XI, 1989 methodology: Edition. Examinations of Category B-A and B-D welds were performed to ASME Section XI Appendix VIII, 1995 Edition up to and including the 1996 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 Two 10-Year inservice inspections have been performed.

inspections:

Number of There were six indications identified in the beltline region during the most recent indications found: inservice inspection. These indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. None of these indications are within the inner 1/1 Oth or 1 inch of the reactor vessel thickness and all are inherently acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a (Reference 7).

Proposed inspection The third inservice inspection is scheduled for 2016. This inspection will be performed in schedule for balance 2026, plus or minus one refueling cycle. The proposed inspection date is consistent with of plant life: the latest revised implementation plan OG-10-238 (Reference 5), plus or minus one refueling outage.

E2-6

Enclosure 2 Proposed Alternative VEGP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR SO.SSa{a){3){i)

Table 3 Details of the Through Wall Cracking Frequency Calculation for VEGP - Unit 2 at 57 EFPY Inputs Reactor Coolant System Temperature, T cl°F]: N/A T wall [inches]: 8.781 CU(1) Ni(1) R.G. CF(1) Fluence Region and Component Material (1 ) 19 2 1\10. 1.99 RTNOT(u) [10 Neutron/cm ,

Description Heat No. [wt%] [wt%] Pos. (1) [OF]

[OF] E> 1.0 MeV]

1 Intermediate Shell Plate R4-1 C-3527-1 0.07 0.63 1.1 44.0 10 3.19 2 Intermediate Shell Plate R4-2 C-3527-2 0.06 0.61 1.1 37.0 10 3.19 3 Intermediate Shell Plate R4-3 C-3552-1 0.05 0.60 1.1 31.0 30 3.19 4 Lower Shell Plate 88825-1 C-3500-1 0.06 0.62 1.1 37.0 40 3.19 5 Lower Shell Plate R8-1 C-4304-1 0.07 0.63 1.1 44.0 40 3.19 6 Lower Shell Plate 88628-1 C-3500-2 0.05 0.59 1.1 31.0 50 3.19 7 Inter. Shell Long. Weld 101-124A 87005 0.05 0.15 2.1 20.7 -10 1.68 8 Inter. Shell Long. Weld 101-1248 87005 0.05 0.15 2.1 20.7 -10 3.07 9 Inter. Shell Long. Weld 101-124C 87005 0.05 0.15 2.1 20.7 -10 3.07 10 Lower Shell Long. Weld 101-142A 87005 0.05 0.15 2.1 20.7 -10 1.68 11 Lower Shell Long. Weld 101-1428 87005 0.05 0.15 2.1 20.7 -10 3.07 12 Lower Shell Long. Weld 101-142C 87005 0.05 0.15 2.1 20.7 -10 3.07 Inter. To Lower Shell Circ. Weld 13 87005 0.05 0.15 2.1 20.7 -30 3.19 101-171 Outputs Methodology Used to Calculate l::..T30: I Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Fluence FF Material RT MAX*XX 19 2 l::..T3O

[OR] [10 Neutron/cm , (Fluence TWCF 9S 'XX Region No. [OF]

E> 1.0 MeV] Factor)

(From Above)

Limiting Axial Weld - AW 5 556.69 3.07 1.296 57.02 O.OOE+OO Limiting Plate - PL 5 557.10 3.19 1.305 57.43 4.87E-14 Circumferential Weld - CW 5 557.10 3.19 1.305 57.43 O.OOE+OO TWCF9S.TOTAL(OAWTWCF9S.AW + oPLTWCF9s.PL + oCWTWCF 9S.CW ): 1.22E-13 Note:

(1) Reference 9 E2-7

Enclosure 2 Proposed Alternative VEGP-ISI-ALT-06 Version 1.0, in Accordance with 10 CFR 50.55a(a)(3)(i)

Table 4 List of Affected Components for VEGP - Unit 2 ASME ASME ITEM COMPONENT 10 DESCRIPTION CATEGORY NUMBER i B-A B1.11 -V6-001-W04 UPPER SHELL TO INTERMEDIATE SHELL WELD B-A B1.11 21201-V6-001-W05 ~EDIATE SHELL TO LOWER SHELL WELD B-A B1.11 21201-V6-001-W06 R SHELL TO BOTTOM HEAD TORUS WELD B-A B1.11 21201-V6-001-W207 NOZZLE DROP-OUT TO SHELL WELD B-A B1.11 21201-V6-001-W208 NOZZLE DROP OUT TO SHELL WELD B-A B1.12 21201-V6-001-W12 UPPER SHELL LONGITUDINAL WELD AT 42 DEGREES B-A B1.12 21201-V6-001-W13 UPPER SHELL LONGITUDINAL WELD AT 162 DEGREES B-A B1.12 21201-V6-001-W14 UPPER SHELL LONGITUDINAL WELD AT 282 DEGREES B-A B1.12 21201-V6-001-W15 INTERMEDIATE SHELL LONGITUDINAL WELD AT 0 DEGREES B-A B1.12 21201-V6-001-W16 INTERMEDIATE SHELL LONGITUDINAL WELD AT 120 DEGREE B-A B1.12 21201-V6-001-W17 INTERMEDIATE SHELL LONGITUDINAL WELD AT 240 DEGREE B~GITUDINALWELDAT90DEGREES B-A R SHELL LONGITUDINAL WELD AT 210 DEGREES B-A 2 GITUDINAL WELD AT 330 DEGREES B- B1.21 21201-V6-001-W01 CLOSURE HEAD DOME TO TORUS WELD I B-A B1.21 21201-V6-001-W07 BOTTOM HEAD TORUS TO BOTTOM HEAD DOME WELD B-A B1.22 21201-V6-001-W08 CLOSURE HEAD TORUS MERIDIONAL WELD AT 45 DE~

B-A B1.22 21201-V6-001-W09 CLOSURE HEAD TORUS MERIDIONAL WELD AT 135 DE B-A B1.22 21201.V6.001.~E HEAD TORUS MERIDIONAL WELD AT 225 DEGREES B-A B1.22 21201 TORUS MERIDIONAL WELD AT 315 DEGREES t= B-A B-A B1.22 B1.22 21201 21201-V6-001-W22 BOTTOM HEAD TORUS MERIDIONAL WELD AT 0 DEGREES BOTTOM HEAD TORUS MERIDIONAL WELD AT 90 DEGREES B-A B1.22 21201-V6-001-W23 BOTTOM HEAD TORUS MERIDIONAL WELD AT 180 DEGREES B-A B1.22 21201-V6-001-W24 BOTTOM HEAD TORUS MERIDIONAL WELD AT 270 DEGREES B-A B1.30 21201-V6-0~EL FLANGE TO SHELL WELD B-A B1.40 21201-V6-0 URE HEAD TORUS TO FLANGE WELD B-D B3.90 21201-V6-001-W25 VESSEL TO OUTLET NOZZLE N1 WELD B-D B3.90 21201-V6-001-W26 VESSEL TO INLET NOZZLE N2 WELD B-D B3.90 21201-V6-001-W27 VESSEL TO INLET NOZZLE N3 WELD B-D B3.90 21201-V6-001-W2B VESSEL TO OUTLET NOZZLE N4 WELD B-D B3.90 21201-V6-001-W29 VESSEL TO OUTLET NOZZLE N5 WELD B-D B3.90 21201-V6-001-W30 VESSEL TO INLET NOZZLE N6 WELD B-D B3.90 21201-V6-001-W31 VESSEL TO INLET NOZZLE N7 WELD B-D B3.90 21201-V6-001-W32 VESSEL TO OUTLET NOZZLE NB WELD B-D B3~-V6-001-IR01 OUTLET NOZZLE N1 INNER RADIUS B-D B3. 1-V6-00HR02 INLET NOZZLE N2 INNER RADIUS B-D B3.100 21201-V6-001-IR03 INLET NOZZLE N3 INNER RADIUS B-D B3.100 21201-V6-001-IR04 OUTLET NOZZLE N4 INNER RADIUS B-D B3.100 21201-V6-001-IR05 OUTLET NOZZLE N5 INNER RADIUS B-D B3.100 21201-V6-001-1 R06 INLET NOZZLE N61NNER RADIUS B-D B3.100 21201-V6-001-IR07 INLET NOZZLE N71NNER RADIUS B-D B3.100 21201-V6-001-1 ROB OUTLET NOZZLE N8 INNER RADIUS

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