ML18047A039

Proposed Alternative to Reactor Vessel Inservice Inspection Intervals (Relief Request RR-ENG-3-14)

ML18047A039
Person / Time
Site:	South Texas
Issue date:	02/15/2018
From:	Page M South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-18003547, RR-ENG-3-14
Download: ML18047A039 (13)

Text

Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, Texas 77483 February 15,2018 NOC-AE-18003547 10CFR50.55a U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket No. STN 50-498 and STN 50-499 Proposed Alternative to Reactor Vessel Inservice Inspection Intervals (Relief Request RR-ENG-3-14)

In accordance with the provisions of 10 CFR 50.55a(z)(1), STP Nuclear Operating Company (STPNOC) requests approval for South Texas Project (STP) Units 1 and 2 to extend the Reactor Vessel (RV) Third Inservice Inspection (ISI) Interval of the RV welds from 2020 (Units 1 and 2) to 2026 (Unit 1) and 2027 (Unit 2).

STPNOC proposes to implement an alternative to the requirement ofASME Section Xl IWB-2412, Inspection Program B, that volumetric examination of RV Examination Categories B-A and B-D be performed once each 10-year ISI interval. The current third ISI interval ends September 24, 2020 and October 18, 2020, respectively. STPNOC proposes to perform the third ASME Section Xl Category B-A and B-D examinations in the fourth ISI interval no later than 2026 and 2027, respectively.

STPNOC requests NRC review and approval of this alternative request by February 2019, to support the use of the proposed alternative.

The enclosed Relief Request RR-ENG-3-14 provides the basis and supporting information for the proposed alternative.

There are no commitments in this letter.

If there are any questions, please contact Craig Younger at 361-972-8186, or Kyle Wallis at 361-972-4687.

lichael Rage General Manager of Engineering rjg

Enclosure:

Proposed Alternative to ASME Section Xl Reactor Vessel Inservice Inspection Intervals (Relief Request RR-ENG-3-14)

STI:34596186

ec:

(paper copy)

Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 Lisa M. Regner Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North (08H04) 11555RockvillePike Rockville, MD 20852 NRC Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 289, Mail Code: MN116 Wadsworth,TX 77483 NOC-AE-18003547 Page 2 of 2 (electronic copy)

Morgan, Lewis & Bockius LLP Paul Bessette U.S. Nuclear Regulatory Commission Lisa M. Regner NRG South Texas LP Kevin Malcarney Jim von Suskil Skip Zahn CPS Energy Kevin Polio Cris Eugster L. D. Blaylock City of Austin Elaina Ball John Wester Texas Dept. of State Health Services Robert Free

NOC-AE-18003547 Enclosure Enclosure Proposed Alternative to ASME Section Xl Reactor Vessel Inservice Inspection Intervals (Relief Request RR-ENG-3-14)

NOC-AE-18003547 Enclosure Page 1 of 10 Proposed Alternative to ASME Section Xl Reactor Vessel Inservice Inspection Intervals (Relief Request RR-ENG-3-14)

A. ASME Code Component(s) Affected The affected components are the South Texas Project (STP) Unit 1 and Unit 2 reactor vessels (RV),

specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section Xl, 2004 Edition (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."

Description Shell Welds Circumferential Shell Welds Longitudinal Shell Welds Head Welds Circumferential Head Welds Meridional Head Welds Shell-to-Flange Weld Nozzle-to-Vessel Welds 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.")

B. ApDlicable ASME Code Edition and Addenda ASME Code Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2004 Edition.

Examination Category B-A B-A B-A B-A B-A B-A B-A B-D B-D Item No.

B1.10 B1.11 B1.12 B1.20 B1.21 B1.22 B1.30 B3.90 B3.100 Table 1 -Applicable ASME Code Edition and Addenda Unit Unit 1 Unit 2 Unit 1 Unit 2 Interval 3

3 4

4 Edition 2004 2004 To Be Determined To Be Determined Start September 25, 2010 October 19, 2010 September 25, 2020 October 19, 2020 End September 24, 2020 October 18, 2020 August 20, 2027 December 15, 2028 C. Applicable ASME Code Reauirement IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor vessel pressure retaining welds identified in Table IWB-2500-1 once each 10-year interval. The STP Unit 1 and STP Unit 2 third 10-year inservice inspection (IS!) interval is scheduled to end on September 24, 2020 and October 18, 2020, respectively. The applicable Code for the fourth 10-year ISI interval will be selected in accordance with the requirements of 10 CFR 50.55a.

NOC-AE-18003547 Enclosure Page 2 of 10 D. Reason for Relief from Code Requirements 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-rem exposure and examination costs.

E. Proposed Alternative and Basis for Use:

South Texas Project Nuclear Operating Company (STPNOC) proposes not to perform the ASME Code required volumetric examination of the STP Unit 1 and STP Unit 2 reactor vessels subject examinations for the third inservice inspection, currently scheduled for 2020 and 2019, respectively.

STPNOC will perform the third ASME Code required volumetric examination of the STP Unit 1 and Unit 2 reactor vessels subject examinations in the fourth inservice inspection interval no later than 2026 and 2027, respectively. The proposed inspection dates for STP Unit 1 and Unit 2 are a deviation from implementation plan presented in OG-10-238 (Reference 2). For Unit 1, the impact to the implementation plan in OG-10-238 would increase the number of inspections in 2026 from two to three, and decrease the number of inspections in 2029 from five to four. For Unit 2, the impact to the implementation plan in OG-10-238 would increase the number of inspections in 2027 from seven to eight, and decrease the number of inspections in 2030 from five to four. Based on Figures 3 and 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the inspection plan and the distribution of inspections over time.

In accordance with 10 CFR 50.55a(z)(1), 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-lnformed 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 STP Unit 1 and 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. The parameters for STP Unit 1 and Unit 2 are bounded by the results of the Westinghouse pilot plant and qualifies STP Unit 1 and Unit 2 for ISI interval extensions.

NOC-AE-18003547 Enclosure Page 3 of 10 Table 1a below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of STP Unit 1. Tables 2a and 3a provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

Table 1a: Critical Parameters for the Application of Bounding Analysis for STP Unit 1 Parameter Dominant Pressurized Thermal Shock (PTS)

Transients in the NRC PTS Risk Study are Applicable Through-Wall Cracking Frequency (TWCF)

Frequency and Severity of Design Basis Transients Cladding Layers (Single/Multiple)

Pilot Plant Basis NRC PTS Risk Study (Reference 5) 1.76E-08 Events per year (Reference 4) 7 heatup/cooldown cycles per year (Reference 4)

Single Layer (Reference AL PIant-Specific Basis PTS Generalization Study (Reference 6) 8.27E-16 Events per year (Calculated per Reference 4)

Bounded by 7 heatup/cooldown cycles per year Single Layer Additional Evaluation Required?

No No No No

NOC-AE-18003547 Enclosure Page 4 of 10 Table 2a below provides a summary of the latest reactor vessel inspection for STP Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the STP Unit 1 reactor vessel.

Table 2a: Additional Information Pertaining to Reactor Vessel Inspection for STP Unit 1 Inspection methodology:

Number of past inspections:

Number of indications found:

Proposed inspection schedule for balance of plant life:

The latest ISI for STP Unit 1 was conducted in accordance with the ASME Code, Section Xl and Section V, 1989 Edition, with no Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi).

Examinations of Category B-A and B-D welds were performed to ASME Section Xl, Appendix VIII, 1995 Edition with 1996 Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future inservice inspections will be performed to ASME Section Xl, Appendix VIII methodology.

Two 10-year inservice inspections have been performed.

One indication identified in the beltline region of the RV was recorded during the last ISI. This subsurface indication is located in the intermediate to lower shell circumferential weld seam (Item 9 in Table 3a), and is allowable per Table IWB-3510-1 of Section Xl of the ASME Code. This indication is not within the inner 1,10th or 1 inch of the reactor vessel thickness.

Therefore, it is inherently acceptable per the requirements of the Alternate Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61 a (Reference 7).

The third inservice inspection is currently scheduled for 2020. This inspection will be performed no later than 2026 refueling outage. 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 2026 from two to three, and decrease the number of inspections in 2029 from five to four. 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.

NOC-AE-18003547 Enclosure Page 5 of 10 Table 3a summarizes the inputs and outputs for the calculation ofthrough-wall cracking frequency for STP Unit 1.

Table 3a: Details of TWCF Calculation for STP Unit 1 at 34 Effective Full Power Years (EFPY)

Inputs Reactor C'oolant System Temperature. Tc [°F]: N/A No.

1 2

3 4

5 6

7 8

9 Region and C'omponeut Description Intermediate Shell Plate Intermediate Shell Plate Inteimediate Shell Plate Lower Shell Plate Lower Shell Plate Lower Shell Plate Intermediate Shell Longitudinal Weld Seams Lower Shell Longitudinal "Weld Seams Liteimediate to Lower SheU Circumferenti.tl Weld Seam Material ID Rl606-1 Rl606-2 Rl606-3 R1622-1 R1622-2 R1632-3 101-124A,B,C 101-142.A,B,C 101-171 Material Heat No.

B-8120-2 B-8120-1 C-4326-2 B-9566-2 B-9575-2 B-9575-1 Heat #89476, Lmde 0091 Flux, Lot #

0145 Heat #89476, Linde0091

Flux, Lot #0145 Heat # 89476.

Liade 124

Flux, Lot #1061 Cu

[wt%]

0.04 0.04 0.05 0.05 0.07 0.05 0-022 0-022 0.022 Tma [inches]:

Ni

[wt%]

0.63 0.61 0.62 0.61 0.64 0.66 0.071 0.071 0.071 R.G. 1^9 Pos.

1.1 2.1 1.1 1.1 1.1 1.1 2.1 2-1 2.1 CF [°F]

26.0 29.6 31.0 31.0 44.0 31.0 31.2 31.2 31.2 KTfaiKyi m

10 0

10

-30

-30

-30

-50

-50

-70 9.10 (mensured)

Fluence (n/crn, E > 1.0 MeVl<l>

2.51E+19 2.51E+19 2.51E+-19 2.51E+19 2.51E+19 2.5IE+19 2.51E+19 2.51E+19 2.51E+19 Outputs Methodology Used to Calculate ATjo: Regulatory Guide 1.99, Revision 2 (Reference 8)

Limiting Axial Weld - AW Limiting Plate - PL Limitmg Ciicumferential Wetd - CW Controlling Mntwial Rezion No.

(From Above) 3 3

3 ctec 2.5 2.5 2.5 KTMAX-XX

[°R]

508 sos 508 Flueuce

[u/cm2, E

>1.0 MeV]

2.51E+19 2.51E+19 2.51E+19 FF (FIuence Factor) 1.2472 1.2472 1.2472 AT,o I°F]

38.66 38.66 3S.66 TWCF,,.TOTAL=(ctAwTWCTgs.Aw + si^TWCS^ + ncwTWCF,,,<w):

TO'CF^xx O.OOOE+OO 3.308E-16 O.OOOE+00 8.27E-K5 (1) Fluence values based on plant-specific analysis of record, WCAP-17482-NP (Reference 9)

NOC-AE-18003547 Enclosure Page 6 of 10 Table 1b below lists the critical parameters investigated in the WCAP and compares the results of the Westinghouse pilot plant to those of STP Unit 2. Tables 2b and 3b provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

Table 1b: Critical Parameters for the Application of Bounding Analysis for STP Unit 2 Parameter Dominant Pressurized Thermal Shock (PTS)

Transients in the NRC PTS Risk Study are Applicable Through-Wall Cracking Frequency (TWCF)

Frequency and Severity of Design Basis Transients Cladding Layers (Single/Multiple)

Pilot Plant Basis NRC PTS Risk Study (Reference 5) 1.76E-08 Events per year (Reference 4) 7 heatup/cooldown cycles per year (Reference 4)

Single Layer (Reference

^_

Plant-Specific Basis PTS Generalization Study (Reference 6) 1.09E-16 Events per year (Calculated per Reference 4)

Bounded by 7 heatup/cooldown cycles per year Single Layer Additional Evaluation Required?

No No No No

NOC-AE-18003547 Enclosure Page 7 of 10 Table 2b below provides a summary of the latest reactor vessel inspection for STP Unit 2 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the STP Unit 2 reactor vessel.

Table 2b: Additional Information Pertaining to Reactor Vessel Inspection for STP Unit 2 Inspection methodology:

Number of past inspections:

Number of indications found:

Proposed inspection schedule for balance of plant life:

The latest ISI for STP Unit 2 was conducted in accordance with the ASME Code, Section Xl and Section V, 1989 Edition, with no Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi).

Examinations of Category B-A and B-D welds were performed to ASME Section Xl, Appendix VIII, 1995 Edition with 1996 Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future inservice inspections will be performed to ASME Section Xl, Appendix VIII methodology.

Two 10-year inservice inspections have been performed.

One indication identified in the beltline region of the RV was recorded during the last ISI. This subsurface indication is located in a lower longitudinal weld seam (Item 8 in Table 3b), and is allowable per Table IWB-3510-1 of Section Xl of the ASME Code. This indication is not within the inner 1,10th or 1 inch of the reactor vessel thickness. Therefore, it is inherently acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a (Reference 7).

The third inservice inspection is currently scheduled for 2019. This inspection will be performed no later than 2027 refueling outage. 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 2027 from seven to eight, and decrease the number of inspections in 2030 from five to four. 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.

NOC-AE-18003547 Enclosure Page 8 of 10 Table 3b summarizes the inputs and outputs for the calculation ofthrough-wall cracking frequency for STP Unit 2.

Table 3b: Details of TWCF Calculation for STP Unit 2 at 34 Effective Full Power Years (EFPY)

Inputs Reactor Cooliint System Temperature, Tc [°F]; N/A No.

1 2

3 4

5 6

7 s

9 Region and Component Descripdun Intermediate Shell Plate lateimediate Shell Plate Intermediate Shell Plate Lower Shell Plate Lower Shell Plate Lower SheU Plate Ititemiedmte SheU Longitudinal Weld Seams Lower Shell Longitudinal Weld Seams Intemiediate to Lower Shell Circumfereatial Weld Seam Material ID R2507-1 R2507-2 R2S07-3 R3022-1 R3022-2 R3022-3 101-124A, B,C 101-142A, B,C 101-171 Material Heat No.

NR 62067-1 NR 62230-1 NR 62248-1 NR. 64647-1 NR 64627-1 NR 64445-1 Heat #

90209. Linde 0091 Flux, Lot # 1054 Heat #

90209, Linde 124 Flux, Lot S 1061 Heat H 90209, Linde 124 Flux, Lot

1. 1061 Cu

[wt%]

0.04 0.05 0.05 0.03 0.04 0.04 0.044 0.044 0.044 Tiran [mches]:

Ni

[wt%]

0.65 0.64 0.61 0.63 0.61 0.60 0.126 0.126 0.126 R.G. 1.9.9 Pos.

2.1 1-1 1.1 1.1 1.1 1.1 1.1 1.1 1.1 CT[°F]

33.4 31.0 31.0 20.0 26.0 26.0 37.8 37.8 37.8 RTsDT(B) t'F]

-10

-10

-40

-30

-40

-40

-70

-70

-70 9.10 (measured)

FIuence (a/cm, E > 1.0 MeV]a>

2.50E+19 2.50E+19 2.50E+19 2.50E+19 2-SOE+I9 2.SOE+19 2.50E+19 2-50E+I9 2.50E+I9 Outputs Methodology Used to Calculate AT,,: Regulatoiy Guide 1.99, Revision 2 CReference 8)

Limiting Axial Weld - AW Limiting Plate - PL Limiting Circumferential Weld - CW Conh.'ollmg Matwiai Region No. CFrom Above) 1 1

1 "XX 2.5 2.5 2.5 RTiux-xx

[°R]

491 491 491 Fluence

[D/cm2, E

>1.0 MeV]

2.50E+19 2.50E+19 2.50E+19 FF (FIueuce Factor) 1-2462 1.2462 1.2462 ATM fF]

41.62 41.62 41.62 TWCFiB^TAt=(uAwTWCF^^+ apiJWCFsi^i. + acwTWCFoi-cw):

TWCF,s,xx O.OOOE+00 4.365E-17 O.OOOE+00 I.09E-I6 (1) Fluence values based on plant-specific analysis of record, WCAP-17636-NP (Reference 10)

F. Duration of Proposed Alternative This request is applicable to the STP Unit 1 and Unit 2 inservice inspection programs for the third and fourth 10-year inspection intervals.

NOC-AE-18003547 Enclosure Page 9 of 10 G. Precedents Relief from this examination requirement to apply the proposed alternative at the South Texas Project Unit 1 and Unit 2 was previously approved by the NRC for the following (with ADAMS Accession No. references):

1. "Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAG Nos. ME8573 and ME8574)," dated April 30, 2013, Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140.

2. "Vogtle Electric Generating Plant, Units 1 and 2 - Request for Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT-06 (TAG Nos. MF2596 and MF2597)," dated March 20, 2014, ADAMS Accession Number ML14030A570.

3. "Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAG Nos. MF1922 and MF1923)," dated March 26, 2014, ADAMS Accession Number ML14079A546.

4. "Sequoyah Nuclear Plant, Units 1 and 2 - Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval (TAG Nos. MF2900 and MF2901)," dated August 1, 2014, ADAMS Accession Number ML14188B920.

5. "Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAG No. MF3596)," dated December 10, 2014, ADAMS Accession Number ML14303A506.

6. "Wolf Creek Generating Station - Request for Relief Nos. 13R-08 and 13R-09 for the Third 10-Year Inservice Inspection Program Interval (TAG Nos. MF3321 and MF3322)," dated December 10, 2014, ADAMS Accession Number ML14321A864.

7. "Callaway Plant, Unit 1 - Request for Relief I3R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAG No.

MF3876)," dated February 10, 2015, ADAMS Accession Number ML15035A148.

8. "Braidwood Station, Units 1 and 2 - Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)

(CAC Nos. MF8191 and MF8192)," dated March 15, 2017, ADAMS Accession Number ML17054C255.

NOC-AE-18003547 Enclosure Page 10 of 10 H. References

1. ASME Boiler and Pressure Vessel Code, Section Xl, 2004 Edition, Rules for Inservice Inspection of Nuclear Power Plants, American Society of Mechanical Engineers, New York.

2. PWROG Letter OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval perWCAP-16168-NP, Revision 1, "Risk-lnformed 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-lnformed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. Nuclear Regulatory Commission, November 2002, (ADAMS Accession Number ML023240437).

4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, "Risk-lnformed 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, (ADAMS Accession No. ML15222A848).

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. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

U.S. Nuclear Regulatory Commission, May 1988, (ADAMS Accession No. ML003740284).

9. Westinghouse Report, WCAP-17482-NP, Revision 0, "Analysis of Capsule W from the South Texas Project Nuclear Operating Company South Texas Unit 1 Reactor Vessel Radiation Surveillance Program," May 11, 2012.

10. Westinghouse Report, WCAP-17636-NP, Revision 0, "Analysis of Capsule W from the South Texas Project Nuclear Operating Company South Texas Unit 2 Reactor Vessel Radiation Surveillance Program," October 22, 2012.