CNL-13-108, Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval
| ML13291A006 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 10/15/2013 |
| From: | James Shea Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CNL-13-108 | |
| Download: ML13291A006 (14) | |
Text
Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-13-108 October 15, 2013 10 CFR 50.4 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328
Subject:
Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval
Reference:
Letter from NRC to TVA, "Issuance of Amendments (TAC Nos. M85308 and M85309) (TS 92-16)," dated March 15, 1994 [ML013330167]
In accordance with Title 10 of the Code of Federal Regulations (10 CFR), Part 50.55a, "Codes and Standards," paragraph (a)(3)(i), Tennessee Valley Authority (TVA) proposes an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," as applicable to Sequoyah Nuclear Plant (SQN), Units 1 and 2. The Code of Record for the current third 10-year interval for SQN, Units 1 and 2, is the ASME Section Xl B&PV Code, 2001 Edition with Addenda through 2003.
WVA is submitting Requests for Alternatives (RFAs) 13-1S1-1 and 13-ISI-2 for Nuclear Regulatory Commission (NRC) approval of a proposed alternative to the requirement of ASME Code, Section Xl, Paragraph IWB-2412, "Inspection Program B," requiring volumetric examination of essentially 100 percent of reactor pressure-retaining welds identified in Table IWB-2500-1 once'each 10-year interval. The proposed alternative is to extend this examination 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) from 10 years to 20 years. TVA has concluded that the proposed alternative provides an acceptable level of quality and safety, in accordance with 10 CFR 50.55a(a)(3)(i). The requisite supporting information 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 Commission Page 2 October 15, 2013 As documented in the NRC's safety evaluation for the referenced license amendments, TVA had previously committed to performing augmented ISI examinations of the reactor pressure vessel nozzles. The provided RFAs do not change these commitments. The frequency for future examinations will be aligned consistent with the proposed 20-year ISI interval for the described Examination Category B-A and B-D welds.
TVA requests approval of the RFAs by June 1, 2014 to support refueling outage planning milestones. Both SQN units are currently in their third ISI Program interval, which extends from June 1, 2006 to April 30, 2016. The next examinations of the Examination Category B-A and B-D welds are scheduled for 2015.
These RFAs are similar to recent alternatives granted for Duke Energy's McGuire Nuclear Station, Unit 2, by letter dated September 6, 2012 [ADAMS Accession No. ML12249A175], and Dominion's Surry Power Station, Units 1 and 2, by letter dated April 30, 2013 [ADAMS Accession No. ML13106A140].
There are no new regulatory commitments contained in this submittal. If you have any questions about this request, please contact Clyde Mackaman at (423) 751-2834.
Respecfly J. W. Shea Vice President, Nuclear Licensing
Enclosures:
- 1. Request for Alternative 13-181-1
- 2. Request for Alternative 13-ISI-2 cc (Enclosures):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 THIRD 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 following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section Xl (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-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" Examination Category Item No.
Description B-A B13.11 Circumferential Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-A B1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section (Throughout this request the above examination categories are referred to as "the subject examinations" 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 essentially 100 percent (%) of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. The Sequoyah, Unit 1, third 10-year inservice inspection (ISI) interval began on June 1, 2006 and is scheduled to end on April 30, 2016. The applicable Code for the fourth 10-year ISI interval will be selected in accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a.
4. Reason for Request
An alternative is requested from the requirement of IWB-2412, "Inspection Program B," that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each 10-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-roentgen equivalent man (man-rem) exposure and examination costs.
E1-1 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 THIRD 10-YEAR INTERVAL 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 required volumetric examination of the Sequoyah, Unit 1, reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the third inservice inspection, currently scheduled for 2015. TVA will perform the third ASME Code required volumetric examination of the Sequoyah, Unit 1, reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in the fourth inservice inspection interval in 2024 plus or minus one refueling outage, in accordance with the applicable Code for the fourth 10-year ISI interval. The applicable Code for the fourth 10-year ISI interval will be selected in accordance with the requirements of 10 CFR 50.55a. The proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2). The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2024, and decrease the number of inspections in 2015 from six to five. Based on Figures 3 and 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.
In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).
The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 4). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Sequoyah, Unit 1, were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for Sequoyah, Unit 1, are bounded by the results of the Westinghouse pilot plant qualifies Sequoyah, Unit 1, for an ISI interval extension. Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of Sequoyah, Unit 1. Tables 2 and 3 provide additional information 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 1 Additional Pilot Plant Evaluation Parameter 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 No Basis Transients cycles per year heatup/cooldown cycles per (Reference 4) year Cladding Layers (Single/Multiple)
Single Layer Single Layer No (Reference 4)
SingleLayer E1-2 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-11-1 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 satisfactory examinations 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, Section Xl, 1995 Edition, with the 1996 Addenda. Examinations of Category B-A and B-D welds were performed to ASME Section Xl, Appendix VIII, 1992 Edition with Addenda through 1993. Future inservice inspections will be 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 most found:
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 Section Xl 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 acceptable per 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 will schedule for balance of be performed in 2024 plus or minus one refueling outage. The proposed plant life:
inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2). The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2024, and decrease the number of inspections in 2015 from six to five. Based on Figures 3 and 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.
- Note 1: Indication 1 is 1.40 inches in length, 0.25 inches in through-wall extent (2a dimension), and is embedded with an 'S' dimension of 2.20 inches. Indication 2 is 0.90 inches in length, 0.30 inches in through-wall extent (2a dimension), and is embedded with an 'S' dimension of 1.90 inches.
E1-3 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-11-1 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).
Table 3:
Details of TWCF Calculation for Sequoyah, Unit 1, at 52 Effective Full-Power Years (EFPY)
Inputs IS & LS Twa11 [inches]:
8.623 Reactor Coolant System Temperature, Tc [°F]:
N/A US Twa,1 [inches]:
10.985 Region and C(
F1)1 Fluence [1019 No.
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 0.0584 Forging 282758 2
Intermediate Shell 980807/
0.15 0.86 1.1 115.6 40 2.66 (IS) Forging 281489 3
Lower Shell (LS) 980919/
0.13 0.76 2.1 109.3 73 2.66 Forging 281587 4
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 0.0584 6
IS to LS Circ. Weld 25295 0.35 0.11 2.1 139.3
-40 2.65 7
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 FF Material 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-10 Limiting 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 third and fourth 10-year inspection intervals.
E1-4 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 THIRD 10-YEAR INTERVAL 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, and ME1636)," dated February 22, 2010 (ADAMS Accession Number ML100290415)
"Safety Evaluation of Relief Requests to Extend the Inservice Inspection Interval for Reactor Vessel Examinations for Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos. ME1478, ME1479, ME1480 and ME1481)," dated February 22, 2010 (ADAMS Accession Number ML100491550)
"Arkansas Nuclear One, Unit 2 - Request for Alternative ANO2-1SI-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)
"Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of the Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME3010)," dated July 12, 2010 (ADAMS Accession Number ML101750402)
"Three Mile Island Nuclear Station, Unit 1 (TMI-1) - Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations, Proposed Alternative Request Nos. RR-09-01 and RR-09-02 (TAC Nos. ME2483 and ME2484),"
dated September 21, 2010 (ADAMS Accession Number ML102390018)
"Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, 2013 (ADAMS Accession Number ML13106A140)
"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 ME7330)," dated September 6, 2012 (ADAMS Accession Number ML12249A175)
E1-5 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 1 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-11-1
- 8. References
- 1. ASME Boiler and Pressure Vessel Code, Section Xl,.2001 Edition through 2003 Addenda, American Society of Mechanical Engineers, New York
- 2. OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.' PA-MSC-0120," July 12, 2010 (ADAMS Accession Number ML11153A033)
- 3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. Nuclear Regulatory Commission, November 2002
- 4. Westinghouse Report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," October 2011 (ADAMS Accession Number ML113060207)
- 5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),"
U.S. Nuclear Regulatory Commission, March 2010
- 6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS)
Risk Results to Additional Plants," U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482)
- 7. 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. Westinghouse Report WCAP-1 7539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," March 2012 (ADAMS Accession Number ML13032A253)
- 9. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988 E1-6 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 THIRD 10-YEAR INTERVAL 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 following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section Xl (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-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" Examination Category Item No.
Description B-A B1.11 Circumferential Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.22 Meridional Head Welds B-A B1.30 Shell-to-Flange Weld B-A B1.40 Head-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section (Throughout this request the above examination categories are referred to as "the subject examinations" 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 essentially 100 percent (%) of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. The Sequoyah, Unit 2, third 10-year inservice inspection (ISI) interval began on June 1, 2006 and is scheduled to end on April 30, 2016. The applicable Code for the fourth 10-year ISI interval will be selected in accordance with the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a.
4. Reason for Request
An alternative is requested from the requirement of IWB-2412, "Inspection Program B," that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each 10-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-roentgen equivalent man (man-rem) exposure and examination costs.
E2-1 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 THIRD 10-YEAR INTERVAL 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 required volumetric examination of the Sequoyah, Unit 2, reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the third inservice inspection, currently scheduled for 2015. TVA will perform the third ASME Code required volumetric examination of the Sequoyah Unit 2 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in the fourth inservice inspection interval in 2024 plus or minus one refueling outage in accordance with the applicable Code for the fourth 10-year ISI interval. The applicable Code for the fourth 10-year ISI interval will be selected in accordance with the requirements of 10 CFR 50.55a. The proposed inspection date is consistent with the latest revised implementation plan, OG-10-238 (Reference 2).
In accordance with 10 CFR 50.55a(a)(3)(i), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).
The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 4). This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Sequoyah, Unit 2, were compared to those obtained from the Westinghouse pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for Sequoyah, Unit 2, are bounded by the results of the Westinghouse pilot plant qualifies Sequoyah, Unit 2, for an ISI interval extension. Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of Sequoyah, Unit 2. Tables 2 and 3 provide additional information 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 2 Additional Pilot Plant Plant-Specific Evaluation Parameter 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 No Basis Transients cycles per year heatup/cooldown (Reference 4) cycles per year Cladding Layers (Single/Multiple)
Single Layer Single Layer No (Reference 4)
SingleLayer E2-2 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-ISI-2 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 satisfactory examinations 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,Section XI, 1995 Edition, with the 1996 Addenda. Examinations of Category B-A and B-D welds were performed to ASME Section Xl, Appendix VIII, 1992 Edition with Addenda through 1993. Future inservice inspections will be 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 zero indications identified in the beltline region during the most found:
recent inservice inspection; therefore, the inservice inspection results for Sequoyah, Unit 2, inherently satisfy the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).
Proposed inspection The third inservice inspection is scheduled for 2015. This inspection will schedule for balance of be performed in 2024 plus or minus one refueling outage. The proposed plant life:
inspection date is consistent with the latest revised implementation plan OG-10-238 (Reference 2).
E2-3 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-ISI-2 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).
Table 3:
Details of TWCF Calculation for Sequoyah, Unit 2, at 52 Effective Full-Power Years (EFPY)
Inputs IS & LS Tw 2i, [inches]:
8.623 Reactor Coolant System Temperature, Tc [IF]:
N/A US Twal [inches]:
10.985 Region and C(
N()Fluence
[1019 Regionand 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
0.0552 Forging 285849 Intermediate 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 2.57 Forging 293323 4
Bottom Head Ring 981177/
0.16 0.77 1.1 122.3 5
0.316 Botto HeadRing 288872 5
US to IS Circ. Weld 721858 0.08 1.0 1.1 108.0 10 0.0552 6
IS to LS Circ. Weld 4278 0.12 0.11 2.1 78.9
-4 2.55 7
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 FF Material Region RTMAx-XX 2
AT3 0 TWCF9 Neutron/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-13 Limiting 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 third and 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, and ME1636)," dated February 22, 2010 (ADAMS Accession Number ML100290415)
E2-4 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-ISI-2 "Safety Evaluation of Relief Requests to Extend the Inservice Inspection Interval for Reactor Vessel Examinations for Salem Nuclear Generating Station, Unit Nos. 1 and 2 (TAC Nos.
ME1478, ME1479, ME1480 and ME1481)," dated February 22, 2010 (ADAMS Accession Number ML100491550)
" "Arkansas Nuclear One, Unit 2 - Request for Alternative ANO2-1SI-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)
" "Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of the Reactor Vessel Inservice Inspection Date to the Year 2020 (Plus or Minus One Outage)
(TAC No. ME301 0)," dated July 12, 2010 (ADAMS Accession Number ML101750402)
"Three Mile Island Nuclear Station, Unit 1 (TMI-1) - Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations, Proposed Alternative Request Nos. RR-09-01 and RR-09-02 (TAC Nos. ME2483 and ME2484)," dated September 21, 2010 (ADAMS Accession Number ML102390018)
"Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, 2013 (ADAMS Accession Number ML13106A140)
"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)
- 8. References
- 1. ASME Boiler and Pressure Vessel Code, Section Xl, 2001 Edition through 2003 Addenda, American Society of Mechanical Engineers, New York
- 2. 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-1 6168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," October 2011 (ADAMS Accession Number ML113060207)
- 5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),"
U.S. Nuclear Regulatory Commission, March 2010 E2-5 of 6 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT, UNIT 2 THIRD 10-YEAR INTERVAL REQUEST FOR ALTERNATIVE 13-ISI-2
- 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 Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D.C., Federal Register, Volume 75, No. 1, dated January 4, 2010 and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010
- 8. Westinghouse Report WCAP-17539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," March 2012 (ADAMS Accession Number ML13032A253)
- 9. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988 E2-6 of 6