AEP-NRC-2018-82, Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination

Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination

ML18331A163
Person / Time
Site:	Cook
Issue date:	11/20/2018
From:	Lies Q American Electric Power, Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2018-82
Download: ML18331A163 (4)

Text

---

;c-.--

m INDIANA . lndian11 Michigan Power MICHIGAN. Cook* Nuclear Plant POWER~** . One CcioH>iace *

• Bridgma'n, Ml 49iOB A unit ofAmerican Electric Power lndianaMichiganPoWer.com November 20, 2018 . AEP-NRC-2018-82 10 CFR 50.55a Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555~0001 Donald C. Cook Nuclear Plant Units 1 and 2 Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination

References:

1. Letter from Q. S. Lies, Indiana Michigan Power Company (l&M), to U.S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, Proposed Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination," dated June 14, 2018, Agencywide Documents Access and Management System Accession No. ML18169A148. .

2. E-mail from A W. Dietrich, NRC, to H. L. Levendosky, l&M, "D.C. Cook Units 1 and 2 - RAI for RPV Threads in Flange Alternative (EPID L-2018-LLR-0084);" dated November 8, 2018.

This letter provides Indiana Michigan Power Company's (l&M), 'licensee forDonald C. Cook Nuclear Plant (CNP) Units 1 and 2, response*to the Request for Additional Information (RAI) by the U. S.

Nuclear Regulatory. Commission (NRC) regarding .an alternative request to the examination requirements of American Society of Mechanical Engineers Boiler and PressureVessel Code (ASME Code), Section XI, Examination Category B-G-1, "Pressure Retaining Bolting, Greaterfhan 2 inches (50 rnm) in Diameter;" Item Number 86.40, Threads in Flange.

• By Reference 1, l&M submitted a request to eliminate the volumetric examination requirements of Section XI of the ASME Code threads in the reactor pressure vessel flange for the remainder of the

• CNP fourth 10-year inservice inspection interval. By Reference 2; the NRC transmitted an RAI concerning the alternative request submitted by l&M in Reference 1.

• l&M is* providing, as an enclosure to this letter, its response to the RAI contained in Referem::e 2. .

U. S. Nuclear Regulatory Commission* **.* AEP-NRt-2018-82 Page 2 There are no new regulatory commitments made in this letter. Should you have any questions, please

• contact Mr. Michael K. Scarpello, Regulatory Affairs Director, at (269) 466-2649.

Sincerely, J~Jf-~

Q. Shane Lies Site Vice President BMC/kmh

Enclosure:

Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination c: R. J. Ancona - MPSC R. F. Kuntz, NRG, Washington D.C.

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

Enclosure to AEP.;.NRC-2018.:.82 .

Response to Request for Additional Information Regarding the Alternative Request for Elimination of the Reactor Pressure Vessel Threads in. Flange Examination

• By letter dated June 14, 2018, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18169A148) (Reference. 1), Indiana Michigan Power Company (l&M),

licensee for Donald C. Cook Nuclear Plant Units 1 and 2, submitted.an alternative request to* the examination requirements of American Society of Mechanical Engineers. Boiler and Pressure Vessel Code (ASME Code), Section XI, Examination Category B-G-1, "Pressure Retaining 13olting, Greater thari 2 inches (50 mm) in Diameter," Item Number B6.40, Threads in Fl,;mge.

The U. S. Nuclear Regulatory Commission (NRC) staff is currently reviewing the submittal, and has determined that _additional information is *needed* in order to complete the review. The*

requests for additional information (RAl)s and l&M's responses are provided below. *

• .r RAl-1.

In the lic_ensee's submittal, under "Flaw Tolerance Evaluation," the licensee states that a linear elastic fracture mechanics (LEFM) evaluation consistent with Subsection /WB-3500 of the ASME Code of Record was performed. Based on the NRG staff's evaluation, it appears that the licensee's LEFM evaluation was performed consistent with ASME Code Section XI, Appendix G, and Section 6.2.2 of EPRI Report No. 3002010~54, dated December 2017. *

- ~ . . . .

Confirm that the licensee's LEFM evaluation was performed consistent with ASME Code Section XI, Appendix G, or provide specificsrelated to how the LEFM evaluation was performed.

l&M Response to RAl-1 l&M confirms that the LEFM evaluation in Reference 1 was performed consistent with Appendix G

. of ASME Code Section XI (Reference 2) and Section 6.2.2 :Of EPRI Report No. 3002010354,

• ... "Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements," (Reference 3) ..

• .

• RAl-2 In the licensee's submittal, under "Flaw Tolerance Evaluation," the licensee* states that the allowable stress intensity factor (K) for preload is 27. 5 kslvin (kilopound per square inch square root inches), based on an ASME Code derived K,c (lower bound fracture toughness) of 53.9 ksi./in and asafety factor of 2. NRG staff calculated an allowable K of 27. O ksi./in based on an ASME Code derived K,c of 53. 9 ksi./in and a safety factor of 2.

• Explain the calculation of allowable K for preload.

. i I

I

~

I

. Enclosure to AEP-NRC-2018-82 Page 2 l&M. Response to RAl-2 a

• • . in developing the response to this request, l&M performed review of EPRI Report 3002010354. *

• (Reference 3), which formed the basis for the calculation of allowable K for* preload. Based on

• discussion .with *the principaf investigators of the EPRI Report, the fracture toughness for the

. flange at (T-RT NDT) at 0°F was calculated correctly as 53.9 ksi"in., but an error was found in the

. application of the safety factor to .the fr~cture toughness of the flange. *This error was carried over into Reference 1. The allowable stress intensity factor at preload should be 27.0 ksi"in. This

• value remains greater than the maximum K values for preload conditions summarized in Table 2.

• of the enclosure to Reference* 1. * *

References:

1. Letter*trom Q. S. Lies, Indiana Michigan Power Company, to U.S. Nuclear Regulatory .

Commission, "Donald C. Cook Nuclear Plant Unit 1* and Unit 2,

• Proposed Alternative:

Request for Elimination of the Reactor Pressure Vessel Threads in Flange Examination,"

dated June 14, 2018, ADAMS Accession No. ML18169A148.

2. ASME Boiler and Pressure Vessel Code, Section XI, 2004 Edition, no Addenda.

3. Reactor Pressure Vessel (RPV) Threads in Flange Examination Requirements, EPRI, Palo Alto, CA: 2017. 3002010354.

• i

- .:** ,*