L-17-299, Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BA-01)

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Request to Extend Certain Reactor Vessel Inspections from 10 to 20 Years (Request 1-TYP-4-BA-01)
ML17297A318
Person / Time
Site: Beaver Valley
Issue date: 10/24/2017
From: Bologna R
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
1-TYP-4-BA-01, L-17-299
Download: ML17297A318 (9)
v • d • e


Text

FENOC' Beaver Valley Power Station P.O. Box 4 Shippingport, PA 15077 FirstEnergy Nuclear Operating Company Richard D. Bologna 724-682-5234 Site Vice President Fax: 724-643-8069 October 24, 2017 10 CFR 50.55a(z)(1)

L-17-299 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Request to Extend Certain Reactor Vessel Inspections From 10 to 20 Years (Request 1-TYP-4-BA-01)

In accordance with the provisions of 10 CFR 50.55a(z), FirstEnergy Nuclear Operating Company (FENOC) hereby requests Nuclear Regulatory Commission (NRG) approval of a proposed alternative to American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME BPV Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," paragraph IWB-2412, "Inspection Program B," and Table IWB-2500-1, "Examination Categories," at Beaver Valley Power Station, Unit No. 1.

ASME BPV Code,Section XI, Paragraph IWB-2412, "Inspection Program B," requires volumetric examination of essentially 100 percent of the total number of reactor vessel pressure-retaining welds identified in Table IWB-2500-1, once each 10-year interval.

The proposed alternative would extend the inservice inspection interval from 10 to 20 years for certain reactor vessel welds. A more detailed description of the proposed alternative and supporting information are presented in the enclosure.

FENOC requests approval of the proposed alternative by March 30, 2018.

Beaver Valley Power Station, Unit No. 1 L-17-299 Page 2 There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A Lentz, Manager- Fleet Licensing, at 330-315-6810.

Richard D. Bologna

Enclosure:

Beaver Valley Power Station, Unit No. 1, 10 CFR 50.55a Request 1-TYP-4-BA-0 1, Revision 0 cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP Site BRP/DEP Representative

Enclosure L-17-299 Beaver Valley Power Station, Unit No. 1, 10 CFR 50.55a Request 1-TYP-4-BA-01, Revision 0 (6 Pages Follow)

Beaver Valley Power Station, Unit No. 1 10 CFR 50.55a Request Number: 1-TYP-4-BA-01, Revision 0 Page 1 of 6 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 components are the Beaver Valley Power Station, Unit 1 (BVPS-1) reactor vessel pressure-retaining welds and full penetration nozzle welds listed below. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV)

Code,Section XI, examination categories and item numbers are also listed. These examination categories and item numbers are from Subarticle IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.

Examination Category Item No. Description B-A 81 .11 Circumferential Shell Welds B-A 81.12 Longitudinal Shell Welds B-A B1 .21 Circumferential Head Welds 8-A 81.22 Meridional Head Welds B-A 81.30 Shell-to-Flange Weld 8-D 83.90 Nozzle-to-Vessel Welds B-O 83.100 Nozzle Inside Radius Section (Throughout this request, the above examination categories are referred to as "the subject examinations")

2. Applicable Code Edition and Addenda

ASME BPV Code,Section XI, 2001 Edition with the 2003 Addenda.

3. Applicable Code Requirement

ASME BPV Code,Section XI, Paragraph IWB-2412, "Inspection Program B," requires volumetric examination of essentially 100 percent of the total number of reactor vessel pressure-retaining welds identified in Table IWB-2500-1, once each 10-year interval.

4. Reason for Request

An alternative is requested to the Paragraph IWB-2412 requirement that volumetric examination of essentially 100 percent of the total number of reactor vessel pressure retaining Examination Category 8-A and B-O welds be performed once each 10-year interval. Extension of the interval between examinations of Category B-A and B-O 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.

Beaver Valley Power Station, Unit No. 1 10 CFR 50.55a Request Number: 1-TYP-4-BA-01, Revision 0 Page 2 of 6

5. Proposed Alternative and Basis for Use As an alternative to ASME BPV Code,Section XI, Paragraph IWB-2412, FirstEnergy Nuclear Operating Company (FENOC) proposes to extend the fourth 10-year inservice inspection interval for BVPS-1 reactor vessel pressure-retaining Examination Category B-A welds and nozzle-to-vessel and nozzle inner radius section Examination Category B-D welds from August 28, 2018 to August 28, 2028. FENOC plans to perform the ASME BPV Code required volumetric examination of the BVPS-1 reactor vessel pressure-retaining Examination Category B-A welds and the nozzle-to-vessel and nozzle inner radius section Examination Category B-D welds in 2027, since there is no BVPS-1 refueling outage currently scheduled in 2028. The proposed inspection date of 2027 is consistent with the date reflected in the latest implementation plan, Pressurized Water Reactor Owners Group (PWROG) Letter OG-10-238 (Reference 1).

In accordance with 10 CFR 50.55a(z)(1 ), an alternate inspection interval is requested based on an acceptably small change in risk by satisfying the risk criteria specified in Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.

The methodology used to conduct this analysis is based on the study defined in Westinghouse topical report WCAP-16168-NP-A, Revision 3 (the WCAP), "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 2).

This study focuses on risk assessments of materials within the beltline region of the reactor vessel wall. Appendix A of the WCAP identifies the parameters to be used to determine if the pilot plant evaluations documented in the WCAP bound a plant specific application. The parameters include:

  • Through Wall Cracking Frequency (TWCF),
  • Frequency and Severity of Design Basis Transients, and
  • Cladding Layers (single/multiple).

BVPS-1 was the Westinghouse pilot plant used in the WCAP, and according to the safety evaluation report for the WCAP, "these plants need not confirm the applicability of the parameters in Appendix A of the TR [topical report] for the current license term."

The current license term referred to in the TR has expired, and therefore, the applicability of parameters in Appendix A of the TR were re-evaluated as follows. The pilot plant information provided in the WCAP for BVPS-1 continues to be applicable for the Dominant PTS Transients in the NRC PTS Risk Study parameter and the Cladding Layers parameter. Therefore, no further evaluation of these two parameters is required for BVPS-1.

Beaver Valley Power Station, Unit No. 1 10 CFR 50.55a Request Number: 1-TYP-4-BA-01, Revision 0 Page 3 of 6 The other two parameters (TWCF parameter, and Frequency and Severity of Design Basis Transients parameter) have been re-evaluated since the pilot plant analysis was performed, as these parameters are time dependent. A review of the latest in service inspection report for BVPS-1 was also conducted, as it was not included in the pilot plant analysis. The results of these revised evaluations and review are discussed below.

Table 1A summarizes the inputs and Table 1B summarizes the outputs for the revised TWCF evaluation for BVPS-1. The revised TWCF value of 1.42E-09 events per year presented in Table 1Bis bounded by the applicable pilot plant TWCF value (1.76E-08 events per year) presented in the WCAP.

Table 1A: DetaHs of TWCF Calculation Inputs for 8VPS-1 at 50 Effective Full Power Years (EFPY) 1, 2 Inter. & Lower Shell Twan [inches]: 8.03125 Upper Shell Twa11 [inches]: 9.15625 Region and R.G. Fluence Material Material Heat Cu Ni CF RTNDT(U)

No. Component 1.99 [OF] [Of] [n/cm 2 ,E >

ID No. [wt%] [wt%]

Description Pos. 1.0 MeV]

Upper Shell 1 86604 123V339VA1 0.12 0.68 1.1 84.2 40 7.18E+18 Forging Intermediate Shell 2 86607-1 C4381-1 0.14 0.62 1,1 100.5 26.8 5.88E+19 Plate Intermediate Shell 3 86607-2 C4381-2 0.14 0.62 1.1 100.5 53.6 5.88E+19 Plate 4 Lower Shell Plate 86903-1 C6317-1 0.21 0.54 1.1 147.2 13,1 5.89E+19 5 Lower Shell Plate 87203-2 C6293-2 0.14 0.57 1.1 98.7 0.4 5.89E+19 305414 (3951 & 0.34 0.61 2.1 216.9 -56 7.18E+18 Upper Shell to 3958) 6 Intermediate Shell 10-714 AOFJ 0.03 0.93 1.1 41.0 10 7.18E+18 Girth Weld FOIJ 0.03 0.94 1.1 41.0 10 7.18E+18 EODJ 0.02 1.04 1.1 27.0 10 7.18E+18 HOCJ 0.02 0.93 1.1 27.0 10 7.18E+18 Intermediate Shell 19-714 7 305424 0.28 0.63 1.1 191.7 -56 1.13E+19 Longitudinal Weld A&8 Intermediate To 8 Lower Shell Girth 11-714 90136 0.27 0.07 1.1 124.3 -56 5.88E+19 Weld Lower Shell 20-714 9 305414 0.34 0.61 2.1 216.9 -56 1.14E+19 Longitudinal Weld A&B

Beaver Valley Power Station, Unit No. 1 10 CFR 50.55a Request Number: 1-TYP-4-BA-01, Revision 0 Page 4 of 6 Table 18: Details ofTWCF Calculation Outputs for BVPS-1 at 50 Effective Full Power Years (EFPY) 2 Methodology Used to Calculate b.T 30: Regulatory Guide 1 .99, Revision 2 3 Controlling Fluence FF RTMAX-XX b.T30 Description Material axx [n/cm 2 , (Fluence TWCF95-XX

[OR] [OF]

Region No. E >1.0 MeV] Factor)

Limiting Axial Weld - AW 9 2.4790 629 1.14E+19 1.0366 224.8 2.812E-12 Limiting Plate PL 4 2.1475 684 5.89E+19 1.4333 211.0 6.557E-10 Limiting Forging - FO 1 2.5000 576 7.18E+18 0.9071 76.4 2.635E-13 Limiting Circumferential 4 2.1478 684 5.88E+19 1.4330 210.9 2.931E-12 Weld-CW TWCF9s-TOTAL = (aAwTWCF95-AW + OPLTWCF95-PL + OFOTWCFgs-Fo + acwTWCF9s-cw): 4 1.42E-09 Notes for Tables 1A and 1B:

1. Information in Table 1A, except wall thickness values, is from WCAP-18102-NP, Revision 0, "Beaver Valley Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," June 2017 (Accession Number ML17284A195).
2. The following words are abbreviated in the tables. Intermediate (Inter.), Number (No.), Identification (ID), Copper (Cu), Nickel (Ni), Weight percent (Wt%), Chemistry Factor (CF), degrees Fahrenheit (°F), degrees Rankine (0 R), Newtons per Centimeter squared (n/cm 2), Energy (E), Million electron Volts (MeV), Wall Thickness (Twa11), Material property that characterizes the reactor vessels resistance to fracture initiating from flaws found in various vessel locations (RTMAX-xx), the shift in charpy v-notch transition temperature at the 30 foot-pound energy level produced by irradiation (AT30), and 95 th percentile factor value of TWCF (TWCF9s-xx).
3. Reg Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

May 1988.

4. The a terms used in Table 1Bare determined as shown on page 36 of NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," March 2010.

The seven plant heatup and seven plant cooldown (heatup/cooldown) cycles per year assumed for the Westinghouse pilot plant study in the WCAP are used to demonstrate that the amount of fatigue crack growth considered in the pilot plant analyses is bounding for specific Westinghouse plants. The projected number of RCS heatup/cooldown cycles for 60 years of operation is 175 transients, and the number of RCS Heatup and cooldown design occurrences (139 each) are listed in BVPS-1 UFSAR Table 4.1-10, "Summary of Reactor Coolant System Design Transients." Over the 60 years of plant operation, the number of projected and design RCS heatup/cooldown cycles per year for BVPS-1 are less than the bounding seven heatup/cooldown cycles per year.

Beaver Valley Power Station, Unit No. 1 10 CFR 50.55a Request Number: 1-TYP-4-BA-01, Revision 0 Page 5 of 6 Table 2 below provides a summary of the latest reactor vessel inspection for BVPS-1 and an evaluation of the recorded indications. This information confirms that reactor vessel examinations have been performed with satisfactory results.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection for BVPS-1 The most recent inservice inspection of the Category 8-A and 8-D welds was performed to ASME BPV Code Section XI, 1989 Edition, with no Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, Inspection methodology:

xv, and xvi) (Implementation of Appendix VIII).

Future inservice inspections will be performed to ASME BPV Code Section XI, Appendix VIII requirements.

Three 10-year inservice inspections have been Number of past inspections:

performed.

There were seven indications identified in the beltline region during the most recent inservice inspection. These indications are acceptable per Table IWB-3510-1 of Section XI of the ASME BPV Number of indications Code. None of these indications are within the inner found:

one-tenth or 1 inch of the reactor vessel thickness.

Therefore, no further evaluation is required and these indications are allowable per 10 CFR 50.61a (Reference 3).

Proposed inspection The fourth inservice inspection is required to be schedule for balance of performed in 2018. It is proposed that this plant life: inspection be performed in 2027.

6. Duration of Proposed Alternative The proposed alternative would extend the duration of the fourth 10-year inservice inspection interval for BVPS-1 reactor vessel pressure-retaining Examination Category 8-A welds, and nozzle-to-vessel and nozzle inner radius section Examination Category 8-D welds to August 28, 2028.

Beaver Valley Power Station, Unit No. 1 10 CFR 50.55a Request Number: 1-TYP-4-BA-01, Revision 0 Page 6 of 6

7. Precedent A similar request for Beaver Valley Power Station, Unit 2 has been recently approved to extend the interval between examinations of Category B-A and B-D reactor vessel welds from 10 years to up to 20 years. This request provided information to address the risk-informed criteria set forth in the WCAP. The NRC letter authorizing the alternative is referenced below.

Subject:

"Beaver Valley Power Station, Unit 2 - Relief From the Requirements of the ASME Code (CAC NOS. MF7212 AND MF7217)," Docket No. 50-412, NRC Letter dated December 27, 2016, Accession Number ML16190A133.

In this letter the NRC staff authorized extending the third 10-year ISi interval from August 28, 2018 to August 28, 2028 for reactor vessel pressure-retaining Examination Category B-A welds and nozzle-to-vessel and nozzle inner radius section Examination Category B-D welds.

8. References
1. 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 (Accession Number ML11153A033).
2. Westinghouse Report, WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," October 2011 (Accession Number ML113060207).
3. 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.

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