ML18213A274

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R. E. Ginna, Relief Request ISI-18 to Extend the Reactor Pressure Vessel Inservice Inspection Interval
ML18213A274
Person / Time
Site: Ginna Constellation icon.png
Issue date: 08/01/2018
From: Jim Barstow
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML18213A274 (8)
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Text

Exelon Generation 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.55a August 1, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 R. E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Relief Request ISl-18 to Extend the Reactor Pressure Vessel lnservice Inspection Interval In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1 ), Exelon Generation Company, LLC (Exelon) requests your review and approval of the attached relief request associated with the lnservice Inspection (ISi) Program for the R. E. Ginna Nuclear Power Plant (Ginna). Specifically, this relief request is associated with extending the reactor pressure vessel ISi interval. Exelon is requesting approval of this relief request for the fifth and sixth intervals. Ginna is currently in the fifth 10-year interval, which began on January 1, 201 O and is currently scheduled to end on December 31, 2019.

We request approval of this relief request by August 1, 2019.

Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully, 2J~_,

James Barstow v-Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Relief Request ISl-18 cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, Ginna USNRC Senior Project Manager, Ginna A. L. Peterson, NYSERDA

Attachment Relief Request ISl-18

10 CFR 50.55a Relief Request Revision O (Page 1 of 6)

Request for Relief ISl-18 to Extend the Reactor Pressure Vessel lnservice Inspection Interval In Accordance with 10 CFR 50.55a(z)(1)

1. ASME Code ComponentCsl Aftected The affected component is the A. E. Ginna Nuclear Power Plant (Ginna) Reactor Pressure Vessel (RPV), specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI (Reference 1) examination Categories and Item Numbers covering examinations of the RPV. These examination Categories and Item Numbers are from IWB-2500 and Table IWB-2500-1 of the ASME B&PV Code, Section XI.

Category B-A welds are defined as "Pressure Retaining Welds in Reactor Vessel." Category 8-D welds are defined as "Full Penetration Welded Nozzles in Vessels."

Examination Category Item No. Description B-A 81 .11 Circumferential Shell Welds B-A 81.30 Shell-to-Flange Weld B-A 81.40 Head-to-Flange Weld 8-D 83.90 Nozzle-to-Vessel Welds B-D 83.100 Nozzle Inside Radius Section

2. Applicable Code Edjtjon and Addenda The fifth 10-year interval of the Ginna lnservice Inspection (ISi) Program is based on the ASME B&PV Code, Section XI, 2004 Edition.
3. Applicable Code Regyjrement 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. Ginna's fifth 10-year ISi interval is scheduled to end on December 31, 2019. The applicable ASME Code Edition for the sixth 10-year ISi interval will be selected in accordance with the requirements of 10 CFR 50.55a.
4. Reason for Regyest 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 1O years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

10 CFR 50.55a Relief Request Revision O (Page 2 of 6)

5. prooosed Alternatiye and Basis for Use Exelon proposes to not perform the ASME Code required volumetric examination of the Reactor Pressure Vessel full penetration pressure-retaining Examination Category B-A and 8-D welds for the fifth ISi interval. Exelon will perform the fifth ASME Code required volumetric examination of the Reactor Pressure Vessel full penetration pressure-retaining Examination Category 8-A and B-D welds in the sixth ISi interval in 2029. This inspection would be before the end of the interval on December 31, 2029. The proposed inspection date is a slight deviation from the implementation plan presented in OG-10-238 (Reference 2), since the implementation plan reflects the next inspection being performed in 2031. The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2029 from five to six and decrease the number of inspections in 2031 from one to zero. 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 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 RPV wall. The results of the calculations 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 Ginna are bounded by the results of the Westinghouse pilot plant qualifies Ginna for an ISi interval extension.

Table 1 below lists the critical parameters investigated in WCAP-16168-NP-A and compares the results of the Westinghouse pilot plant to those of Ginna. Tables 2 and 3 provide additional information that was requested by the NRC and is also included in Appendix A of Reference 4.

Table 1: Critical Parameters for the Application of Bounding Analysis for RGE Parameter Pilot Plant Basis Plant-Specific Basis Additional Evaluation Required?

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

Study are Applicable Through-Wall Cracking Frequency 1.76E-08 Events per 3.26E-11 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)

10 CFR 50.55a Relief Request Revision 0 (Page 3 of 6)

Table 2 provides a summary of the latest reactor vessel inspection results for Ginna and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the Ginna Reactor Vessel. Table 3 summarizes the inputs and outputs for the calculation of TWCF.

Table 2: Additional Information Pertaining to the Reactor Vessel Inspections The latest Reactor Pressure Vessel ISi inspection for Ginna was conducted in accordance with the ASME Code, Section XI, 1995 Edition, with 1996 Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv, and xvi) applicable to Inspection the time of the exam. Examinations of Category B-A and B-D welds were methodology: performed to ASME Section XI, 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 XI, Appendix VIII methodology.

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

inspections:

No indications identified in the beltline region of the reactor vessel were Number of indications recorded during the last ISi. Therefore, the result is inherently acceptable per found: the requirements of the Alternate PTS Rule, 10 CFR 50.61 a (Reference 7).

The fifth inservice inspection is currently scheduled for the end of the interval.

With implementation of the extended ISi interval, this inspection will be performed in 2029, before the end of the interval on December 31, 2029. The proposed inspection date is a slight deviation from the latest revised Proposed inspection implementation plan, OG-10-238 (Reference 2), since the implementation schedule for balance of plan reflects the next inspection being performed in 2031 for Ginna. The plant life: impact to the implementation plan in OG-10-238 would increase the number of inspections in 2029 from five to six, and decrease the number of inspections in 2031 from one to zero. Based on Figures 3 and 4 of OG 238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.

10 CFR 50.55a Relief Request Revision O (Page 4 of 6)

Table 3: Details of TWCF Calculation at 53 Effective Full Power Years (EFPY) 1 lnputs Reactor Coolant System Temperature TC [°F]: NIA Twall [inches]: Nozzle Shell 9.156 Intermediate and Lower Shells 6.656 No. Region and Component Material ID Cu [wt%] Ni [wt%] R.G. 1.99 Pos. CF [ 0 F) RTNDT(U) [ 0 F] 2 Fluence [n/cm , E >

Description 1.0 MeV]

1 Nozzle Shell Forging 123P118 0.17 0.68 1.1 129.0 30 2.37E+18 2 Intermediate Shell Forging 125S255 0.07 0.69 1.1 44.0 20 5.56E+19 3 Lower Shell Forging 125P666 0.05 0.69 2.1 46.2 40 5.56E+19 4 Nozzle Shell to Intermediate Shell Linde 80 0.23 0.59 1.1 167.6 10 2.37E+18 Girth Weld (Heat # 71249) 5 Intermediate Shell to Lower Shell Linde 80 0.25 0.56 1.1 170.4 -4.8 5.56E+19 Girth Weld (Heat# 61782)

Outputs Methodology Used to Calculate bT30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Material RTMAX-XX Fluence [n/cm 2, E FF (Fluence Region No. axx [°R] Factor) bT30 [°F) TWCF95.xx

>1.0 MeV]

Limiting Forging - FO 1 2.5000 568.45 2.37E+18 0.6107 78.78 1.359E-13 Limiting Circumferential Weld - CW 5 2.0662 697.29 5.56E+19 1.4227 242.42 1.560E-11 TWCF95-TOTAL = (OFQTWCF95-FO + acwTWCF9s-cw): 3.26E-11 (1) Material properties and fluence values are taken from WCAP-17214-NP (Reference 9).

10 CFR 50.SSa Relief Request Revision O (Page 5 of 6)

6. puration of proposed A!ternatjye This request is applicable to the A. E. Ginna Nuclear Power Plant lnservice Inspection Program for the fifth and sixth 10-year IS! 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, Agencywide Document Access and Management System (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.
  • "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. l3R-08 and l3R-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.

10 CFR 50.55a Relief Request Revision 0 (Page 6 of 6)

8. References
1) ASME Boiler and Pressure Vessel Code, Section XI, 2004 Edition, American Society of Mechanical Engineers, New York.
2) PW ROG Letter OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice Inspection Interval perWCAP-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 ML023240437).
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 ML11306A084).
5) NUREG-187 4, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," U.S. Nuclear Regulatory Commission, March 2010 (ADAMS Accession Number 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 50.61a, "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 Number ML003740284).
9) Westinghouse Report, WCAP-17214-NP, Rev. 0, "R. E. Ginna Heatup and Cooldown Limit Curves for Normal Operation and Pressurized Thermal Shock Evaluation,"

July 2010.

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