RA-20-0104, Response to NRC Request for Additional Information - Catawba Nuclear Station-SG Report (RA-19-0391)-C2R23

Response to NRC Request for Additional Information - Catawba Nuclear Station-SG Report (RA-19-0391)-C2R23

ML20097B706
Person / Time
Site:	Catawba
Issue date:	04/06/2020
From:	Hare M Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-20-0104
Download: ML20097B706 (4)

Text

( ~ DUKE Mandy Hare Nuclear Support Services Manager

~ ENERGY~ Catawba Nuclear Station Duke Energy CN03CH I 4800 Concord Rd York, SC 29745 803. 701.2218 Mandy. Hare@duke-energy.com 10 CFR 50.55 April 6, 2020 Serial: RA-20-0104 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Catawba Nuclear Station, Unit No. 2 Docket No. 50-414 / Renewed License No. NPF-52

SUBJECT:

Response to NRC Request for Additional Information - Catawba Nuclea r

Station - SG Report (RA-19-0391) - C2R23

REFERENCES:

1. Duke Energy Letter, Catawba Unit 2, Refuel (C2R23) lnservice Inspect ion (ISi) and Steam Generator Inspection (SG-ISI) Repo,t (RA-19-0391), dated Decem ber 19, 2019 (ADAMS Accession No. ML19353A416).

2. NRC Email from M. Mahoney to A. Zaremba, Request for Additional Inform ation-Catawba Nuclear Station - SG Repo,t (RA-19-0391) - C2R23, dated March 13, 2020 (ADAMS Accession No. ML20073F383).

Ladies and Gentlemen:

By letter dated December 19, 2019, Duke Energy Carolinas, LLC (Duke Energy) submitted the Catawba Unit 2, Refuel 23 (C2R23) lnservice Inspection (ISi) and Steam Generator lnservice Inspection (SG-ISI) Summary Reports (Reference 1). By the email dated March 13, 2020 (Reference 2), the NRC requested additional information regarding the SG-ISI report. The Duke Energy response to the request for additional information (RAI) is provide d in the Enclosure.

This submittal contains no regulatory commitments. Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Manager - Nuclear Fleet Licensing, at 980-373-2062.

Sincerely, Mandy Hare Nuclear Support Services Manager, Catawba Nuclear Station

U.S. Nucle ar Regulatory Comm ission Serial: RA-20-0104 Page 2 NDE cc:

M. Mahoney, NRC Project Manager, NRR L. Dudes, NRC Regional Administrator, Region II J.D. Austin, NRC Senio r Resid ent Inspector, Cataw ba Nucle ar Station

U.S. Nuclear Regulatory Commission Serial: RA-20-0104 Enclosure, Page 1 Enclo sure Duke Energy Carolinas, LLC Serial: RA-20-0104 Duke Energy Response to NRC Request for Addit ional Infor matio n for the Catawba Unit 2 Steam Gene rator Inspe ction Repo rt (RA-19-0391)

U.S. Nuclear Regulatory Commission Serial: RA-20-0104 Enclosure, Page 2 NRC Request By letter RA-19-0391, dated December 19, 2019 (Agen cywide Documents Access and Management System Accession No. ML19353A416

), Duke Energy (the licensee) submitted information summarizing the results of the fall 2019 steam generator inspections performed at Catawba Nuclear Station, Unit 2. These inspections were performed during refueling outage 23 (RFO 23). A report from each steam generator inspe ction is submitted to the U. S. Nuclear Regulatory (NRC) in accordance with the plant Techn ical Specifications (TS).

The NRC staff has reviewed the application and, based upon this review, determined that additional information is needed to complete our review

. Please provide a response on the docket within 30 days of this correspondence.

Request for Additional Information (RAl-01)

The tube at Row 31 Column 33 (R31C33) in Steam Generator 2A (SG 2A) was plugged during refueling outage 23 (RFO 23) due to an eddy current indication of axial cracking just above the 3H tube support plate. The tube inspection report states that the indication met the condition monitoring criteria for leakage because the eddy curre nt voltage was less than the threshold voltage for leakage. In this case, the measured voltag e was being compared to a screening criterion for axial outside diameter stress corrosion cracking (ODSCC) in the EPRI Steam Generator In-Situ Pressure Test Guidelines. In the tube inspection report, this single axial indication (SAi) lists two voltage measurements of 2.05 V and 2.09 V from Channel P4.

a. For in-situ pressure-test screening, the EPRI Guide lines provide +Point probe threshold voltage values for the 300 kHz analysis frequency.

What was the 300 kHz voltage for this flaw?

b. Please describe the frequency or frequencies used for Channel P4 and how Channel P4 relates to the 300 kHz signal in terms of performing the screening.

c. If the 300 kHz voltage exceeds the threshold value, discuss how this flaw would be processed through the in-situ pressure test guidelines.

Duke Energy Response

a. The 300 kHz voltage for this flaw is 2.09 volts, the same as the P4 channel.

b. P4 is 300 / 100 kHz mix channel to suppress the tube support plate. Ch 3 is the 300 kHz channel. Both were normalized to the requirements of Steam Generator Management Program: Steam Generator In-Situ Press ure Test Guidelines, Revision 5.

EPRI, Palo Alto, CA: 2016 3002007856.

c. The 300 kHz voltage did not exceed the threshold voltag e for leakage.