AEP-NRC-2018-51, Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examination - Relief Request ISIR-4-08

From kanterella
Jump to navigation Jump to search

Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examination - Relief Request ISIR-4-08
ML18215A179
Person / Time
Site: Cook American Electric Power icon.png
Issue date: 08/01/2018
From: Lies Q
Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2018-51, TAC MC9768
Download: ML18215A179 (9)
v • d • e


Text

Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power lndianaMichiganPower.com August 1, 2018 AEP-NRC-2018-51 10 CFR 50.55a Docket No.: 50-315 U.S. Nuclear Regulatory Commission ATTN:. Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Unit 1 Request to Extend the lnservice Inspection Interval for Reactor Vessel Weld Examination -

Relief Request ISIR-4-08

Reference:

Letter from Ho K. Nieh, Nuclear Regulatory Commission, to Gordon Bischoff, Westinghouse Owner's Group, regarding Final Safety Evaluation for PWROG Topical Report WCAP-16168-NP, Revision 2, (TAC NO. MC9768), dated May 8, 2008

-(ML081060051).

Pursuant to 10 CFR 50.55a(z)(1), Indiana Michigan Power" Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1, hereby requests Nuclear Regulatory Commission (NRC) approval of the following request for the fourth 10-year interval inservice inspection testing program.

l&M is requesting approval of Relief Request ISIR-4-08 for use of risk-informed extension of the Reactor Pressure Vessel lnservice Inspection (ISi) Interval from 10 to 20 years. This relief request is applicable to the CNP Unit 1 inservice inspection program for the fourth and fifth 10-year inspection intervals.

The NRC approved WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of The Reactor Vessel In-Service Inspection Interval," in the above referenced letter. This WCAP provides for extension of the ISi interval for certain pressure retaining welds in the reactor vessel from 10 to 20 years. l&M proposes to implement this extended ISi interval for Unit 1. The plant-specific

_ information identified by the above letter needed to support this request is included in the Enclosure to this letter. l&M has concluded that the proposed alternative provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1 ).

l&M requests approval of the relief request by January 31, 2019, to allow use of the alternative during the Spring 2019 Unit 1 refueling outage U1C29. -

U. S. Nuclear Regulatory Commission AEP-NRC-2018-51 Page 2 This letter contains no new or revised commitments. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely, Z~J.t, Q.~ne Lies Sife Vice President JMT/mll

Enclosure:

10 CFR 50.55a Relief Request ISIR-4-08

U.S. Nuclear Regulatory Commission AEP-NRC-2018-51 Page 3 c: R. J. Ancona - MPSC A. W. Dietrich, NRG, Washington, D.C.

MDEQ - RMD/RPS NRG Resident Inspector K. S. West, NRG Region Ill A. J. Williamson - AEP Ft. Wayne, w/o enclosures

10 CFR 50.55a Relief Request ISIR-4-08 Proposed Alternative for Donald C. Cook Unit 1 In Accordance with 10 CFR 50.55a(z)(1)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected Pursuant to 10 CFR 50.55a(z)(1), Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1, hereby requests Nuclear Regulatory Commission (NRC) approval of the following request for the fourth 10-year interval inservice inspection (ISi) testing program. The affected component is the CNP 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 examinatio_n 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-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 CNP Unit 1 fourth 10-year ISi interval is scheduled to end on February 29, 2020.

The applicable Code for the fifth 10-year ISi interval will be selected in accordance with the requirements of 10 CFR 50.55a.

Enclosure to AEP-NRC-2018-51 Page2

4. Reason for Request

An alternative is requested from the requirement of IW8-2412, Inspection Program 8, that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category 8-A and 8-0 welds be performed once each 10-year interval. Extension of the interval between examinations, of Category 8-A and 8-0 welds, from 10 years to up to 20 years will result in a reduction in person-rem exposure and examination costs.

5. Proposed Alternative and Basis for Use l&M proposes not to perform the ASME Code required volumetric examination of the CNP Unit 1 RV full penetration pressure-retaining Examination Category 8-A and 8-0 welds for the fourth inservice inspection, currently scheduled for 2019. l&M will perform the fourth ASME Code required volumetric examination of the CNP Unit 1 RV full penetration pressure-retaining Examination Category 8-A and 8-0 welds in the fifth inservice inspection interval in 2029. The proposed inspection date for CNP Unit 1 is consistent with the implementation plan presented in 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 through analysis of Unit 1 RV that satisfies the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to conduct the analysis of Unit 1 is based on WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval" (Reference 4). WCAP-16168-NP-A, Revision 3, focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for CNP Unit 1 were compared, in the Unit 1 analysis, 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 Unit 1 analysis demonstrates that the parameters for CNP Unit 1 are bounded by the results of the Westinghouse pilot plant, which qualifies CNP Unit 1 for consideration of 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 CNP Unit 1. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

Enclosure to AEP-NRC-2018-51 Page 3 Table 1: Critical Parameters for the Application of Bounding Analysis for CNP Unit 1 Additional Pilot Plant Plant-Specific Parameter Evaluation Basis Basis Required?

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

Risk Study are Applicable 1.27E-09 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 cycles per year heatup/cooldown No Basis Transients (Reference 4) cycles per year Single Layer Cladding Layers (Single/Multiple) Single Layer No (Reference 4)

Table 2 below provides a summary of the latest reactor vessel inspection for CNP Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the CNP Unit 1 reactor vessel.

Table 2: Additional Information Pertaining to Reactor Vessel Inspection for CNP Unit 1 The latest ISi for CNP Unit 1 was conducted in accordance with the ASME Code,Section XI and Section V, 1989 Edition, with no Addenda.

Examinations of Category B-A and B-D welds were performed to ASME Inspection methodology:

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 Three 10-year inservice inspections have been performed.

inspections:

No indications were* identified in the beltline region of the RV during the Number of indications last inservice inspection. Therefore, the inservice inspection results found: inherently satisfy the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).

The fourth inservice inspection is currently scheduled for 2019. This Proposed inspection inspection is proposed to be performed during the Fall 2029 refueling schedule for balance of outage. The proposed inspection date is consistent with the latest revised plant life:

implementation plan, OG-10-238 (Reference 2).

Table 3 below summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Enclosure to AEP-NRC-2018-51 Page4 Table 3: Details of TWCF Calculation for CNP Unit 1 at 48 Effective Full Power Years (EFPY)

Inputs (References 9 and 10)(1)

Upper Shell T wall [inches]: 10.969 Intermediate and Lower Shell Twa11 [inches]: 8.844 Fluence 1" 1 Material Heat Cu Ni R.G. 1.99 RTNDT(U) 2 No. Region and Component Description Material ID CF [°F] [n/cm ,

No. [wt%] [wt%] Pos. [OF]

E > 1.0 MeV]

1 Upper Shell Plate 84405-1 - 0.14 0.46 1.1 93.7 10 4.27E+17 2 Upper Shell Plate 84405-2 - 0.14 0.45 1.1 93.3 26 4.27E+17 3 Upper Shell Plate 84405-3 - 0.14 0.48 1.1 94.6 30 4.27E+17 4 Intermediate Shell Plate 84406-1 - 0.12 0.52 1.1 81.4 5 2.78E+19 5 Intermediate Shell Plate 84406-2 - 0.15 0.50 1.1 104.5 33 2.78E+19 6 Intermediate Shell Plate 84406-3 - 0.15 0.49 1.1 104.0 40 2.78E+19 7 Lower Shell Plate 84407-1 - 0.14 0.55 1.1 97.8 28 2.81 E+19 8 Lower Shell Plate 84407-2 - 0.12 0.59 1.1 82.8 -12 2.81 E+19 9 Lower Shell Plate 84407-3 - 0.14 0.50 1.1 95.5 38 2.81E+19 10 Upper Shell Longitudinal Welds 1-442 A, 8, and C 13253/12008 0.21 0.873 1.1 208.7 -56 3.24E+17 2-442 A, 8, and 11 Intermediate Shell Longitudinal Welds 13253/12008 0.21 0.873 1.1 208.7 -56 1.85E+19 C

12 Lower Shell LonQitudinal Welds 3-442 A, 8, and C 13253/12008 0.21 0.873 1.1 208.7 -56 1.87E+19 Upper to Intermediate Shell 13 8-442 20291 0.216 0.737 1.1 188.4 -56 4.27E+17 Circumferential Weld Intermediate to Lower Shell 14 9-442 1P3571 0.287 0.756 2.1 218.6 -56 2.78E+19 Circumferential Weld Outputs Methodology Used to Calculate Lff30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Fluence FF 2

Controlling Material Region No. axx RT MAX-XX [

0 R] [n/cm , E >1.0 (Fluence .aT30 [°F] TWCFss-xx MeV] Factor)

Limiting Axial Weld -AW 12 2.3611 648.15 1.87E+19 1.1715 244.48 5.157E-10 Limiting Plate - PL 6 2.4580 632.00 2.78E+19 1.2724 132.33 2.106E-11 Limiting Circumferential Weld -

14 2.1591 681.81 2.78E+19 1.2724 278.14 2.263E-12 cw 1WCF95-TOTAL = (aAw1WCF9s-Aw + OpL1WCF9s-PL + Ocw1WCF9s-cw): 1.27E-09 (1) Material properties for the intermediate and lower shell plate and weld materials are based on WCAP-15878 (Reference 9). Material properties for the upper shell plate and weld materials are based on AEP letter, AEP-NRC-2015-94 (Reference 10).

(2) Fluence values based on plant-specific analysis of record.

Enclosure to AEP-NRC-2018-51 Page 5

6. Duration of Proposed Alternative This request is applicable to the CNP Unit 1 ISi program for the fourth and fifth 10-year inspection intervals.
7. Precedents
  • "Donald C. Cook Nuclear Plant, Unit 2 (CNP-2) - Evaluation of Relief Request (ISIR-29) to extend the third 10-year inservice inspection (ISi) interval for reactor vessel weld examination (TAC No. MD9934}," dated June 8, 2009, Agencywide Document Access and Management System (ADAMS) Accession Number ML091260163.
  • "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-ISI-ALT-05 and VEGP-ISI-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. 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 R~lief 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.

Enclosure to AEP-NRC-2018-51 Page 6

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 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-1874, "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 50.61 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. NuGlear Regulatory Commission, May 1988, (ADAMS Accession Number ML003740284).
9. Westinghouse Report, WCAP-15878, Revision 0, "D. C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation for 40 Years and 60 Yec;1rs,"

December 2002.

10. AEP Letter, AEP-NRC-2015-94, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Response to License Condition Regarding Analysis of Pressure and Temperature Curves for Ferritic Reactor Vessel Materials," September 25, 2015, (ADAMS Accession Number ML15272A491).
Retrieved from "https://kanterella.com/w/index.php?title=AEP-NRC-2018-51,_Request_to_Extend_the_Inservice_Inspection_Interval_for_Reactor_Vessel_Weld_Examination_-_Relief_Request_ISIR-4-08&oldid=2070587"