tennessee valley authority, 1101 market street ... · sequoyah, unit 1, reactor vessel full...

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402

CNL-13-108

October 15, 201310 CFR 50.4

10 CFR 50.55a

ATTN: Document Control DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001

Sequoyah Nuclear Plant, Units 1 and 2Facility Operating License Nos. DPR-77 and DPR-79NRC Docket Nos. 50-327 and 50-328

Subject: Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor VesselWeld Inservice Inspection Interval

Reference: Letter from NRC to TVA, "Issuance of Amendments (TAC Nos. M85308 andM85309) (TS 92-16)," dated March 15, 1994 [ML013330167]

In accordance with Title 10 of the Code of Federal Regulations (10 CFR), Part 50.55a, "Codesand Standards," paragraph (a)(3)(i), Tennessee Valley Authority (TVA) proposes an alternativeto the requirements of the American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel (B&PV) Code, Section Xl, "Rules for Inservice Inspection of Nuclear PowerPlant Components," as applicable to Sequoyah Nuclear Plant (SQN), Units 1 and 2. The Codeof Record for the current third 10-year interval for SQN, Units 1 and 2, is the ASME Section XlB&PV Code, 2001 Edition with Addenda through 2003.

WVA is submitting Requests for Alternatives (RFAs) 13-1S1-1 and 13-ISI-2 for NuclearRegulatory Commission (NRC) approval of a proposed alternative to the requirement of ASMECode, Section Xl, Paragraph IWB-2412, "Inspection Program B," requiring volumetricexamination of essentially 100 percent of reactor pressure-retaining welds identified inTable IWB-2500-1 once'each 10-year interval. The proposed alternative is to extend thisexamination frequency to once each 20-year interval.

WVA proposes to apply the alternative to the third and fourth 10-year Inservice Inspection (ISI)Program intervals to extend the interval for the examination of reactor vessel welds(Examination Category B-A) and nozzle-to-vessel welds (Examination Category B-D) from10 years to 20 years. TVA has concluded that the proposed alternative provides an acceptablelevel of quality and safety, in accordance with 10 CFR 50.55a(a)(3)(i). The requisite supportinginformation and basis for use are provided in the enclosed RFAs 13-11-1 and 13-ISI-2 for SQN,Units 1 and 2, respectively.

Printed on recycled paper

U.S. Nuclear Regulatory CommissionOctober 15, 2013

As documented in the NRC's safety evaluation for the referenced license amendments, TVAhad previously committed to performing augmented ISI examinations of the reactor pressurevessel nozzles. The provided RFAs do not change these commitments. The frequency forfuture examinations will be aligned consistent with the proposed 20-year ISI interval for thedescribed Examination Category B-A and B-D welds.

TVA requests approval of the RFAs by June 1, 2014 to support refueling outage planningmilestones. Both SQN units are currently in their third ISI Program interval, which extends fromJune 1, 2006 to April 30, 2016. The next examinations of the Examination Category B-A andB-D welds are scheduled for 2015.

These RFAs are similar to recent alternatives granted for Duke Energy's McGuire NuclearStation, Unit 2, by letter dated September 6, 2012 [ADAMS Accession No. ML12249A1 75], andDominion's Surry Power Station, Units 1 and 2, by letter dated April 30, 2013 [ADAMSAccession No. ML13106A140].

There are no new regulatory commitments contained in this submittal. If you have anyquestions about this request, please contact Clyde Mackaman at (423) 751-2834.

Respecfly

J. W. SheaVice President, Nuclear Licensing

Enclosures:

1. Request for Alternative 13-181-12. Request for Alternative 13-ISI-2

cc (Enclosures):

NRC Regional Administrator - Region IINRC Senior Resident Inspector - Sequoyah Nuclear Plant

Enclosure 1TENNESSEE VALLEY AUTHORITY

SEQUOYAH NUCLEAR PLANT, UNIT 1THIRD 10-YEAR INTERVAL

REQUEST FOR ALTERNATIVE 13-11-1

1. ASME Code Component(s) Affected

The affected component is the Sequoyah, Unit 1, reactor vessel (RV), specifically the followingAmerican Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section Xl (Reference 1) examination categories and item numbers covering examinations ofthe RV. These examination categories and item numbers are from IWB-2500 and TableIWB-2500-1 of the ASME BPV Code, Section Xl.

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"

ExaminationCategory Item No. DescriptionB-A B13.11 Circumferential Shell WeldsB-A B1.21 Circumferential Head WeldsB-A B1.22 Meridional Head WeldsB-A B1.30 Shell-to-Flange WeldB-A B1.40 Head-to-Flange WeldB-D B3.90 Nozzle-to-Vessel WeldsB-D B3.100 Nozzle Inside Radius Section

(Throughout this request the above examination categories are referred to as "the subjectexaminations" and the ASME BPV Code, Section Xl, is referred to as "the Code.")

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"2001 Edition through 2003 Addenda (Reference 1).

3. Applicable Code Requirement

IWB-2412, "Inspection Program B," requires volumetric examination of essentially100 percent (%) of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 onceeach 10-year interval. The Sequoyah, Unit 1, third 10-year inservice inspection (ISI) intervalbegan on June 1, 2006 and is scheduled to end on April 30, 2016. The applicable Code for thefourth 10-year ISI interval will be selected in accordance with the requirements of Title 10 of theCode of Federal Regulations (10 CFR) 50.55a.

4. Reason for Request

An alternative is requested from the requirement of IWB-2412, "Inspection Program B," thatvolumetric examination of essentially 100% of reactor vessel pressure-retaining ExaminationCategory B-A and B-D welds be performed once each 10-year interval. Extension of theinterval between examinations of Category B-A and B-D welds from 10 years to up to 20 yearswill result in a reduction in man-roentgen equivalent man (man-rem) exposure and examinationcosts.

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REQUEST FOR ALTERNATIVE 13-1S1-1

5. Proposed Alternative and Basis for Use

The Tennessee Valley Authority (TVA) proposes to not perform the ASME Code requiredvolumetric examination of the Sequoyah, Unit 1, reactor vessel full penetration pressure-retainingExamination Category B-A and B-D welds for the third inservice inspection, currently scheduledfor 2015. TVA will perform the third ASME Code required volumetric examination of theSequoyah, Unit 1, reactor vessel full penetration pressure-retaining Examination Category B-Aand B-D welds in the fourth inservice inspection interval in 2024 plus or minus one refuelingoutage, in accordance with the applicable Code for the fourth 10-year ISI interval. Theapplicable Code for the fourth 10-year ISI interval will be selected in accordance with therequirements of 10 CFR 50.55a. The proposed inspection date is a deviation from the latestrevised implementation plan, OG-10-238 (Reference 2). The impact to the implementation plan inOG-10-238 would increase the number of inspections in 2024, and decrease the number ofinspections in 2015 from six to five. Based on Figures 3 and 4 of OG-10-238, this proposedinspection schedule is considered to have a minor impact on the future inspection plan and thedistribution of inspections over time.

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on thebasis that the current interval can be revised with negligible change in risk by satisfying the riskcriteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the studyWCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-ServiceInspection Interval" (Reference 4). This study focuses on risk assessments of materials withinthe beltline region of the RV wall. The results of the calculations for Sequoyah, Unit 1, werecompared to those obtained from the Westinghouse pilot plant evaluated inWCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to becompared. Demonstrating that the parameters for Sequoyah, Unit 1, are bounded by the resultsof the Westinghouse pilot plant qualifies Sequoyah, Unit 1, for an ISI interval extension. Table 1below lists the critical parameters investigated in the WCAP and compares the results of theWestinghouse pilot plant to those of Sequoyah, Unit 1. Tables 2 and 3 provide additionalinformation that was requested by the NRC and included in Appendix A of Reference 4.

Table 1: Critical Parameters for the Application of Bounding Analysis for Sequoyah, Unit 1Additional

Pilot Plant EvaluationParameter Basis Plant-Specific Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization Study No(PTS) Transients in the NRC PTS (Reference 5) (Reference 6)Risk Study are Applicable

Through-Wall Cracking Frequency 1.76E-08 Events per 6.38E-10 Events per year No(TWCF) year (Reference 4) (Calculated per Reference 4)Frequency and Severity of Design 7 heatup/cooldown Bounded by 7 NoBasis Transients cycles per year heatup/cooldown cycles per

(Reference 4) year

Cladding Layers (Single/Multiple) Single Layer Single Layer No(Reference 4) SingleLayer

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Table 2 below provides a summary of the latest reactor vessel inspection for Sequoyah, Unit 1,and an evaluation of the recorded indications. This information confirms that satisfactoryexaminations have been performed on the Sequoyah, Unit 1, reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection for Sequoyah, Unit I

Inspection methodology: The latest ISI was conducted in accordance with the ASME Code, SectionXl, 1995 Edition, with the 1996 Addenda. Examinations of Category B-Aand B-D welds were performed to ASME Section Xl, Appendix VIII,1992 Edition with Addenda through 1993. Future inservice inspections willbe performed to ASME Section Xl, Appendix VIII requirements.

Number of past Two 10-year inservice inspections have been performed.inspections:

Number of indications There were two indications identified in the beltline region during the mostfound: recent inservice inspection. These subsurface indications are located in

the intermediate shell to lower shell circumferential weld (Item 6 in Table 3,below). Both indications are acceptable per Table IWB-3510-1 of SectionXl of the ASME Code. None of these indications are within the inner 1 / 1 0 th

or 1 inch of the reactor vessel thickness and all are inherently acceptableper the requirements of the Alternate PTS Rule, 10 CFR 50.61a(Reference 7). *See Note 1.

Proposed inspection The third inservice inspection is scheduled for 2015. This inspection willschedule for balance of be performed in 2024 plus or minus one refueling outage. The proposedplant life: inspection date is a deviation from the latest revised implementation plan,

OG-10-238 (Reference 2). The impact to the implementation plan inOG-10-238 would increase the number of inspections in 2024, anddecrease the number of inspections in 2015 from six to five. Based onFigures 3 and 4 of OG-10-238, this proposed inspection schedule isconsidered to have a minor impact on the future inspection plan and thedistribution of inspections over time.

*Note 1: Indication 1 is 1.40 inches in length, 0.25 inches in through-wall extent (2a dimension), and isembedded with an 'S' dimension of 2.20 inches. Indication 2 is 0.90 inches in length, 0.30 inches inthrough-wall extent (2a dimension), and is embedded with an 'S' dimension of 1.90 inches.

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Table 3 summarizes the inputs and outputs for the calculation of through-wall crackingfrequency (TWCF).

Table 3: Details of TWCF Calculation for Sequoyah, Unit 1, at 52 Effective Full-Power Years (EFPY)Inputs

IS & LS Twa11 [inches]: 8.623Reactor Coolant System Temperature, Tc [°F]: N/A

US Twa,1 [inches]: 10.985

Region and C( F1)1 Fluence [1019No. Component Material Cu Ni(1) R.G. 1.99 CF( 1) RTNDT(u) Neutron/cm2,N. cmptionen Heat No. [wt%] [wt%] Pos.(1 ) [OF] [OF] E> 1.0 MeV]Description E>10MV

1 Upper Shell (US) 980950/ 0.16 0.89 1.1 123.9 23 0.0584Forging 2827582 Intermediate Shell 980807/ 0.15 0.86 1.1 115.6 40 2.66

(IS) Forging 2814893 Lower Shell (LS) 980919/ 0.13 0.76 2.1 109.3 73 2.66

Forging 2815874 Bottom Head Ring 981177/ 0.16 0.77 1.1 122.3 5 0.336

BotomHea Rng 288872 1___5 US to IS Circ. Weld 25006 0.17 1.0 1.1 207.0 10 0.05846 IS to LS Circ. Weld 25295 0.35 0.11 2.1 139.3 -40 2.657 LSto Bottom Head 25295 0.35 0.11 2.1 139.3 -40 0.336

Ring Circ. Weld

Outputs

Methodology Used to Calculate AT30 : Regulatory Guide 1.99, Revision 2(2)

Controlling Fluence [1019 FFMaterial Region RTMA-XX Neutron/cm 2, (Fluence AT 30

No. (From [OR) Ne 1Factor) [OF] TWCFFxx

Above) E > 1.0 MeV] Factor)

Limiting Forging - FO 3 670.56 2.66 1.262 137.89 2.86E-10Limiting Circ. Weld - CW 3 670.46 2.65 1.261 137.79 4.01E-13

TWCF95-TOTAL(aFoTWCF95.FO + acwTWCF 95_cw): 6.38E-1 0(1) Reference 8(2) Reference 9

6. Duration of Proposed Alternative

This request is applicable to the Sequoyah, Unit 1, inservice inspection program for the thirdand fourth 10-year inspection intervals.

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REQUEST FOR ALTERNATIVE 13-1I1-1

7. Precedents

* "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, andME1636)," dated February 22, 2010 (ADAMS Accession Number ML100290415)

* "Safety Evaluation of Relief Requests to Extend the Inservice Inspection Interval forReactor 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)

* "Arkansas Nuclear One, Unit 2 - Request for Alternative ANO2-1SI-004, to Extend theThird 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations (TACNo. ME2508)," dated September 21, 2010 (ADAMS Accession Number ML1 02450654)

* "Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension ofthe Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus OneOutage) (TAC No. ME3010)," dated July 12, 2010 (ADAMS Accession NumberML101750402)

• "Three Mile Island Nuclear Station, Unit 1 (TMI-1) - Request to Extend the InserviceInspection Interval for Reactor Vessel Weld and Internal Examinations, ProposedAlternative Request Nos. RR-09-01 and RR-09-02 (TAC Nos. ME2483 and ME2484),"dated September 21, 2010 (ADAMS Accession Number ML1 02390018)

* "Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor VesselInspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, 2013 (ADAMSAccession Number ML13106A140)

* "McGuire Nuclear Station, Unit 2, Relief 10-MN-002 to Extend the Inservice InspectionInterval for Reactor Vessel Category B-A and B-D Welds (TAC Nos. ME7329 andME7330)," dated September 6, 2012 (ADAMS Accession Number ML1 2249A1 75)

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8. References

1. ASME Boiler and Pressure Vessel Code, Section Xl,.2001 Edition through 2003Addenda, American Society of Mechanical Engineers, New York

2. OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of ExtendedInservice Inspection Interval per WCAP-16168-NP, Revision 1, 'Risk-Informed Extensionof the Reactor Vessel In-Service Inspection Interval.' PA-MSC-0120," July 12, 2010(ADAMS Accession Number ML1 11 53A033)

3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic RiskAssessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis," U.S. Nuclear Regulatory Commission, November 2002

4. Westinghouse Report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of theReactor Vessel In-Service Inspection Interval," October 2011 (ADAMS AccessionNumber ML1 13060207)

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. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate Fracture ToughnessRequirements for Protection Against Pressurized Thermal Shock Events," U.S. NuclearRegulatory Commission, Washington D.C., Federal Register, Volume 75, No. 1, datedJanuary 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-1 7539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," March 2012 (ADAMS AccessionNumber ML13032A253)

9. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor VesselMaterials," U.S. Nuclear Regulatory Commission, May 1988

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REQUEST FOR ALTERNATIVE 13-1SI-2

1. ASME Code Component(s) Affected

The affected component is the Sequoyah, Unit 2, reactor vessel (RV), specifically the followingAmerican Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section Xl (Reference 1) examination categories and item numbers covering examinations ofthe RV. These examination categories and item numbers are from IWB-2500 and TableIWB-2500-1 of the ASME BPV Code, Section Xl.

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"

ExaminationCategory Item No. DescriptionB-A B1.11 Circumferential Shell WeldsB-A B1.21 Circumferential Head WeldsB-A B1.22 Meridional Head WeldsB-A B1.30 Shell-to-Flange WeldB-A B1.40 Head-to-Flange WeldB-D B3.90 Nozzle-to-Vessel WeldsB-D B3.100 Nozzle Inside Radius Section

(Throughout this request the above examination categories are referred to as "the subjectexaminations" and the ASME BPV Code, Section XI, is referred to as "the Code.")

2. Applicable Code Edition and Addenda

ASME Code Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components,"2001 Edition through 2003 Addenda (Reference 1).

3. Applicable Code Requirement

IWB-2412, "Inspection Program B," requires volumetric examination of essentially100 percent (%) of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 onceeach 10-year interval. The Sequoyah, Unit 2, third 10-year inservice inspection (ISI) intervalbegan on June 1, 2006 and is scheduled to end on April 30, 2016. The applicable Code for thefourth 10-year ISI interval will be selected in accordance with the requirements of Title 10 of theCode of Federal Regulations (10 CFR) 50.55a.

4. Reason for Request

An alternative is requested from the requirement of IWB-2412, "Inspection Program B," thatvolumetric examination of essentially 100% of reactor vessel pressure-retaining ExaminationCategory B-A and B-D welds be performed once each 10-year interval. Extension of theinterval between examinations of Category B-A and B-D welds from 10 years to up to 20 yearswill result in a reduction in man-roentgen equivalent man (man-rem) exposure and examinationcosts.

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REQUEST FOR ALTERNATIVE 13-ISI-2

5. Proposed Alternative and Basis for Use

The Tennessee Valley Authority (TVA) proposes to not perform the ASME Code requiredvolumetric examination of the Sequoyah, Unit 2, reactor vessel full penetration pressure-retainingExamination Category B-A and B-D welds for the third inservice inspection, currently scheduledfor 2015. TVA will perform the third ASME Code required volumetric examination of theSequoyah Unit 2 reactor vessel full penetration pressure-retaining Examination Category B-A andB-D welds in the fourth inservice inspection interval in 2024 plus or minus one refueling outage inaccordance with the applicable Code for the fourth 10-year ISI interval. The applicable Code forthe fourth 10-year ISI interval will be selected in accordance with the requirements of10 CFR 50.55a. The proposed inspection date is consistent with the latest revisedimplementation plan, OG-10-238 (Reference 2).

In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on thebasis that the current interval can be revised with negligible change in risk by satisfying the riskcriteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the studyWCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-ServiceInspection Interval" (Reference 4). This study focuses on risk assessments of materials withinthe beltline region of the RV wall. The results of the calculations for Sequoyah, Unit 2, werecompared to those obtained from the Westinghouse pilot plant evaluated inWCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to becompared. Demonstrating that the parameters for Sequoyah, Unit 2, are bounded by the resultsof the Westinghouse pilot plant qualifies Sequoyah, Unit 2, for an ISI interval extension. Table 1below lists the critical parameters investigated in the WCAP and compares the results of theWestinghouse pilot plant to those of Sequoyah, Unit 2. Tables 2 and 3 provide additionalinformation that was requested by the NRC and included in Appendix A of Reference 4.

Table 1: Critical Parameters for the Application of Bounding Analysis for Sequoyah, Unit 2Additional

Pilot Plant Plant-Specific EvaluationParameter Basis Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization No(PTS) Transients in the NRC PTS (Reference 5) Study (Reference 6)Risk Study are Applicable

Through-Wall Cracking Frequency 1.76E-08 Events per 1.27E-12 Events per No(TWCF) year (Reference 4) year (Calculated per

Reference 4)

Frequency and Severity of Design 7 heatup/cooldown Bounded by 7 NoBasis Transients cycles per year heatup/cooldown

(Reference 4) cycles per year

Cladding Layers (Single/Multiple) Single Layer Single Layer No(Reference 4) SingleLayer

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Table 2 below provides a summary of the latest reactor vessel inspection for Sequoyah, Unit 2,and an evaluation of the recorded indications. This information confirms that satisfactoryexaminations have been performed on the Sequoyah, Unit 2, reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection for Sequoyah, Unit 2

Inspection methodology: The latest ISI was conducted in accordance with the ASME Code, SectionXI, 1995 Edition, with the 1996 Addenda. Examinations of Category B-Aand B-D welds were performed to ASME Section Xl, Appendix VIII, 1992Edition with Addenda through 1993. Future inservice inspections will beperformed to ASME Section Xl, Appendix VIII requirements.

Number of past Two 10-Year inservice inspections have been performed.inspections:

Number of indications There were zero indications identified in the beltline region during the mostfound: recent inservice inspection; therefore, the inservice inspection results for

Sequoyah, Unit 2, inherently satisfy the requirements of the Alternate PTSRule, 10 CFR 50.61a (Reference 7).

Proposed inspection The third inservice inspection is scheduled for 2015. This inspection willschedule for balance of be performed in 2024 plus or minus one refueling outage. The proposedplant life: inspection date is consistent with the latest revised implementation plan

OG-10-238 (Reference 2).

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Table 3 summarizes the inputs and outputs for the calculation of through-wall crackingfrequency (TWCF).

Table 3: Details of TWCF Calculation for Sequoyah, Unit 2, at 52 Effective Full-Power Years (EFPY)Inputs

IS & LS Tw2i, [inches]: 8.623Reactor Coolant System Temperature, Tc [IF]: N/A

US Twal [inches]: 10.985

Region and C( N()Fluence [1019Regionand Material Cu Ni~l) R.G. 1.99 CF(1) RTNDT(u)1 Neutron/cm2,

No. Component Heat No. [wt%] [wt%] Pos.() [OF] [OF] Neu>t1.Description Ei >i 1.0 MeV]

1 Upper Shell (US) 981201/ 0.16 0.84 1.1 123.4 5 0.0552Forging 285849Intermediate Shell 288757/

2 IS)Frging 981057 0.13 0.76 2.1 91.1 10 2.57(IS) Forging 981057

3 Lower Shell (LS) 990469/ 0.14 0.76 1.1 104.0 -22 2.57Forging 293323

4 Bottom Head Ring 981177/ 0.16 0.77 1.1 122.3 5 0.316Botto HeadRing 288872 ____

5 US to IS Circ. Weld 721858 0.08 1.0 1.1 108.0 10 0.05526 IS to LS Circ. Weld 4278 0.12 0.11 2.1 78.9 -4 2.557 LStoBottomHead 721858 0.08 1.0 1.1 108.0 10 0.316

Ring Circ. Weld _

Outputs

Methodology Used to Calculate AT30 : Regulatory Guide 1.99, Revision 2(2)

Controlling Fluence [1019 FFMaterial Region RTMAx-XX 2 AT3 0 TWCF9Neutron/cm , (Fluence TWCos-x

No. (From [OR] E>1.0 MeV] Factor) F]Above)

Limiting Forging - FO 2 583.83 2.57 1.253 114.16 5.10E-13Limiting Circ. Weld - CW 2 583.65 2.55 1.251 113.98 0.OOE+00

TWCF9s-TOTAL(0FoTWCF95-FO + acwTWCFg95CW): 1.27E-12

(1) Reference 8(2) Reference 9

6. Duration of Proposed Alternative

This request is applicable to the Sequoyah, Unit 2, inservice inspection program for the thirdand fourth 10-year inspection intervals.

7. Precedents

* "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, andME1636)," dated February 22, 2010 (ADAMS Accession Number ML100290415)

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* "Safety Evaluation of Relief Requests to Extend the Inservice Inspection Interval for ReactorVessel Examinations for Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos.ME1478, ME1479, ME1480 and ME1481)," dated February 22, 2010 (ADAMS AccessionNumber ML100491550)

" "Arkansas Nuclear One, Unit 2 - Request for Alternative ANO2-1SI-004, to Extend the Third10-Year Inservice Inspection Interval for Reactor Vessel Weld Examinations (TAC No.ME2508)," dated September 21, 2010 (ADAMS Accession Number ML1 02450654)

" "Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of theReactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage)(TAC No. ME301 0)," dated July 12, 2010 (ADAMS Accession Number ML101750402)

* "Three Mile Island Nuclear Station, Unit 1 (TMI-1) - Request to Extend the InserviceInspection Interval for Reactor Vessel Weld and Internal Examinations, Proposed AlternativeRequest Nos. RR-09-01 and RR-09-02 (TAC Nos. ME2483 and ME2484)," datedSeptember 21, 2010 (ADAMS Accession Number ML102390018)

* "Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor VesselInspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, 2013 (ADAMSAccession Number ML13106A140)

* "McGuire Nuclear Station, Unit 2, Relief 10-MN-002 to Extend the Inservice InspectionInterval for Reactor Vessel Category B-A and B-D Welds (TAC Nos. ME7329 andME 7330)," dated September 6, 2012 (ADAMS Accession Number ML1 2249A1 75)

8. References

1. ASME Boiler and Pressure Vessel Code, Section Xl, 2001 Edition through 2003Addenda, American Society of Mechanical Engineers, New York

2. OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of ExtendedInservice Inspection Interval per WCAP-16168-NP, Revision 1, 'Risk-Informed Extensionof the Reactor Vessel In-Service Inspection Interval.' PA-MSC-0120," July 12, 2010(ADAMS Accession Number ML1 1153A033)

3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic RiskAssessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis," U.S. Nuclear Regulatory Commission, November 2002

4. Westinghouse Report WCAP-1 6168-NP-A, Revision 3, "Risk-Informed Extension of theReactor Vessel In-Service Inspection Interval," October 2011 (ADAMS AccessionNumber ML1 13060207)

5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),"U.S. Nuclear Regulatory Commission, March 2010

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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. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate Fracture ToughnessRequirements for Protection Against Pressurized Thermal Shock Events," U.S. NuclearRegulatory Commission, Washington D.C., Federal Register, Volume 75, No. 1, datedJanuary 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-17539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," March 2012 (ADAMS AccessionNumber ML13032A253)

9. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor VesselMaterials," U.S. Nuclear Regulatory Commission, May 1988

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tennessee valley authority, 1101 market street ... · sequoyah, unit 1, reactor vessel full...

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