SBK-L-19003, Relief Request Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l), 3IR-17 Implementation of an Extended Reactor Vessel Inservice Inspection Interval

Relief Request Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l), 3IR-17 Implementation of an Extended Reactor Vessel Inservice Inspection Interval

ML19011A331
Person / Time
Site:	Seabrook
Issue date:	01/11/2019
From:	Brown K NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-19003
Download: ML19011A331 (9)

Text

NEXTeraM ENERGy_~

SEABROOK January 11; 2019 10 CFR 50.55a Docket No. 50-443 SBK-L-19003 U.S . Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20582 Seabrook Station Relief Request Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l), 3IR-17 Implementation of an Extended Reactor Vessel Inservice Inspection Interval In accordance with the provisions of 10 CFR 50.55a(z)(l), NextEra Energy Seabrook, LLC (NextEra) requests approval to extend the inservice inspection interval for the Seabrook Unit 1 Reactor Vessel (RV) from 2020 to 2029.

NextEra proposes to implement an alternative to the requirement of ASME Section XI 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 on August 18, 2020. NextEra proposes to perform the third ASME Section XI Category B-A and B-D examinations in the fourth ISI interval no later than 2029. The attachment to this letter provides the basis and supporting information for the proposed alternative.

NextEra requests prompt NRC review and approval of the proposed alternative by January 15, 2020, to support the use of the proposed alternative during the refueling outage in the spring of 2020.

There are no commitments being made in this submittal.

If you have any questions regarding this submittal, please contact me at (603) 773-7932.

Sincerely, NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L-19003/Page 2

Attachment:

Proposed Alternative for Seabrook Unit lin Accordance with 10 CFR 50.55a(z)(l) cc:

NRC Region I Administrator NRC Project Manager NRC Senior Resident Inspector

Attachment to SBK-L-19003 Relief Request Number 3IR-l 7 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l), Implementation of an Extended Reactor Vessel Inservice Inspection Interval

Relief Request Number 31R-17 Extension of Seabrook Unit 1 RPV Welds from 10 to 20 Years Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the Seabrook Unit 1 reactor vessel (RV), specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section XI (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 XI.

Category B-A welds are defined as "Pressure Retaining Welds in Reactor Vessel."

Category B-D welds are defined as "Full Penetration Welded Nozzles in Vessels."

Examination Category Item No. Description B-A 81 .10 Shell Welds B-A 81 .11 Circumferential Shell Welds B-A 81 .12 Longitudinal Shell Welds B-A 81.20 Head Welds B-A 81.21 Circumferential Head Welds B-A 81.22 Meridional Head Welds B-A 81.30 Shell-to-Flange Weld B-A 81.40 Head-to-Flange Weld B-D 83.90 Nozzle-to-Vessel Welds B-D 83.100 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 XI, "Rules for lnservice Inspection of Nuclear Power Plant Components,"

2004 Edition (Reference 1).

3. Applicable Code Requirement

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 Seabrook Unit 1 third 10-year inservice inspection (ISi) interval is scheduled to end on August 18, 2020. The applicable Code for the fourth 10-year ISi interval will be selected in accordance with the requirements of 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-rem exposure and examination costs.

SBK-L-19003 Attachment Page 2of6

5. Proposed Alternative and Basis for Use NextEra proposes not to perform the ASME Code required volumetric examination of the Seabrook Unit 1 RV full penetration pressure-retaining Examination Category B-A and B-D welds for the third inservice inspection, currently scheduled for 2020. NextEra will perform the third ASME Code required volumetric examination of the Seabrook Unit 1 RV full penetration pressure-retaining Examination Category B-A and B-D welds in the fourth inservice inspection interval in 2029. The proposed inspection date is consistent with the latest revised implementation plan in PWROG letter OG-10-238 (Reference 2).

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 in 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 Seabrook 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 Seabrook Unit 1 are bounded by the results of the Westinghouse pilot plant qualifies Seabrook 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 Seabrook 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 Seabrook Unit 1 Additional Pilot Plant Parameter Plant-Specific Basis Evaluation Basis Required?

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

Study are Applicable 1.32E-13 Events per Through-Wall Cracking Frequency 1.76E-08 Events per year (Calculated per No (TWCF) year (Reference 4)

Reference 4) 7 heatup/cooldown Bounded by 7 Frequency and Severity of Design Basis cycles per year heatup/cooldown No Transients (Reference 4) cvcles per vear Single Layer Cladding Layers (Single/Multiple) Single Layer No (Reference 4)

SBK-L-19003 Attachment Page 3of6 Table 2 below provides a summary of the latest reactor vessel inspection for Seabrook Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the Seabrook Unit 1 reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection for Seabrook Unit 1 The latest ISi for Seabrook Unit 1 was conducted in accordance with the ASME Code, Section XI, 1995 Edition, with 1996 Addenda. Examinations of Category 8-A and 8-D welds were performed to ASME Section XI, Appendix VIII, 1995 Edition with 1996 Inspection methodology:

Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi). Future inservice inspections will continue to be performed to ASME Section XI, Appendix VIII methodology.

Number of past Two 10-year inservice inspections and a preservice inspection have been performed.

inspections:

There were 18 indications identified in the beltline region of the RV during the last ISi.

The subsurface indications are located in the upper, intermediate and lower shell axial weld seams, and the intermediate to lower shell circumferential weld seam (Items 10, 12, 14 and 13, respectively, in Table 3), and are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Two of these indications are within the inner 1/1 oth or 1 inch of the reactor vessel thickness and required further evaluation. These indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a (Reference 7).

A disposition of the two flaws against the limits of the Alternate PTS Rule is shown in the table below. Both flaws are located within the plate material of the reactor vessel.

Number of indications Scaled Maximum number of flaws per 8132 found: Through-Wall Number of square-inches of inside surface area in the Extent, TWE (in) Seabrook inspection volume that are greater than or Flaws equal to TWEM1N and less than TWEMAX*

Evaluated TWEMIN TWEMAX This flaw density does not include (Axial/Ci re.)

underclad cracks in forqinqs.

0 0.075 No Limit 0 0.075 0.375 66 2 (1 /1) 0.125 0.375 26 2 (1 /1) 0.175 0.375 7 1 (0/1) 0.225 0.375 3 1 (0/1) 0.275 0.375 1 1 (0/1) 0.325 0.375 1 1 (0/1) 0.375 Infinite 0 0 Proposed inspection The third inservice inspection is currently scheduled for 2020. This schedule for inspection is proposed to be performed during the 2029 refueling balance of outage. The proposed inspection date is consistent with the latest plant life: revised implementation plan, OG-10-238 (Reference 2).

SBK-L-19003 Attachment Page 4of6 Table 3 below summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation for Seabrook Unit 1 at 55 Effective Full Power Years (EFPY)

Inputs (Reference 9)11 l Upper Shell Twan [inches]: 10.91 Intermediate and Lower Shell Twan [inches]: 8.79 Fluence Material Heat Cu Ni R.G.1.99 RTNDT(U)

No. Region and Component Description Material ID CF [°F] [n/cm 2 ,

No. [wt%] [wt%] Pos. [oF]

E > 1.0 MeV]

1 Upper Shell (US) Plate R1807-3, 122-102A C4233-2 0.06 0.67 1.1 37 10 7.35E+17 2 US Plate R1807-1, 122-1028 C4049-1 0.08 0.60 1.1 51 30 7.35E+17 3 US Plate R1807-2, 122-102C C4049-2 0.09 0.61 1.1 58 30 7.35E+17 4 Intermediate Shell (IS) Plate R1806-1, 124-102A C4036-2 0.045 0.61 1.1 28.5 40 3.05E+19 5 IS Plate R1806-3, 124-1028 C4197-1 0.075 0.63 1.1 47.5 10 3.05E+19 6 IS Plate R1806-2, 124-102C A2749-2 0.06 0.64 1.1 37 0 3.05E+19 7 Lower Shell (LS) Plate R1808-3, 142-102A 01136-2 0.07 0.59 2.1 45.0 40 3.05E+19 8 LS Plate R1808-1, 142-1028 01081-3 0.06 0.58 1.1 37 40 3.05E+19 9 LS Plate R1808-2, 142-102C 01081-2 0.06 0.58 1.1 37 10 3.05E+19 10 US Longitudinal (Long) Welds 101-122 A,8,C 86998 0.05 0.11 1.1 38.7 -10 7.12E+17 11 US to IS Circumferential (Circ) Weld 103-121 90128 0.045 0.06 1.1 31.3 -56 7.35E+17 12 IS Long Welds 101-124 A,8,C 4P6052 0.047 0.049 1.1 30.7 -60 1.78E+19 13 IS to LS Circ Weld 101-171 4P6052 0.047 0.049 1.1 30.7 -60 3.03E+19 14 LS Long Welds 101-142 A,8,C 4P6052 0.047 0.049 1.1 30.7 -60 2.93E+19 Outputs Methodology Used to Calculate l\.T 30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Fluence FF Controlling Material Region No. axx RT MAX-XX [°R] [n/cm 2 , E >1.0 (Fluence l\.T30 [°F] TWCFss.xx MeV] Factor)

Limiting Axial Weld - AW 7 2.5000 557.50 2.93E+19 1.2851 57.83 O.OOOE+OO Limiting Plate - PL 7 2.5000 557.93 3.05E+19 1.2946 58.26 5.261E-14 Limiting Circumferential Weld - CW 7 2.5000 557.86 3.03E+19 1.2931 58.19 O.OOOE+OO TWCFgs.TQTAL =(aAwTWCFgs..Aw + OpLTWCFgs.PL + acwTWCFgs.cw): 1.32E-13 (1) Material properties and fluence values for the upper, intermediate and lower shell plate and weld materials are based on WCAP-17441-NP (Reference 9).

SBK-L-19003 Attachment Page 5of6

6. Duration of Proposed Alternative This request is applicable to the Seabrook Unit 1 inservice inspection program for the third and fourth 10-year inspection intervals.

7. Precedents

• "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.

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

• "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 (TAC Nos. MF1922 and MF1923)," dated March 26, 2014, ADAMS Accession Number ML14079A546.

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

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

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

• "Callaway Plant, Unit 1 - Request for Relief 13R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAC No. MF3876)," dated February 10, 2015, ADAMS Accession Number ML15035A148.

• "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.

• "South Texas Project, Units 1 and 2 - Relief from the Requirements of the ASME Code Regarding the Third 10-Year lnservice Inspection Program Interval (EPID L-2018-LLR-0010)," dated July 24, 2018, ADAMS Accession Number ML18177A425.

SBK-L-19003 Attachment Page 6of6

8. References

1. ASME Boiler and Pressure Vessel Code, Section XI, 2004 Edition, American Society of Mechanical Engineers, New York.

2. PWROG Letter OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice 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, (ADAMS Accession Number ML003740133).

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, (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, 201 O 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-17441-NP, Revision 0, "Seabrook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," October 2011, (ADAMS Accession No. ML12341A096).