ML17172A084

Request for Additional Information Regarding Relief Request to Utilize ASME Code Case N-702

ML17172A084
Person / Time
Site:	Nine Mile Point
Issue date:	06/27/2017
From:	Marshall M Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co
Marshall M, DORL/LPL1, 415-2871
References
CAC MF9381, CAC MF9382
Download: ML17172A084 (3)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 27, 2017 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST TO UTILIZE ASME CODE CASE N-702 (CAC NOS. MF9381 AND MF9382)

Dear Mr. Hanson:

By letter dated March 7, 2017 (Agencywide Documents Access and Management System Accession No. ML17067A056), Exelon Generation Company, LLC submitted Relief Request NMP-RR-001 for the Nine Mile Point Nuclear Station, Units 1 and 2. The proposed relief request would authorize an alternative to performing 100 percent examination of the reactor pressure vessel nozzles listed in the request. The proposed alternative is to examine a minimum of 25 percent of the nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1."

The U.S. Nuclear Regulatory Commission (NRG) staff has reviewed the information provided in the March 7, 2017, letter and has determined that additional information is needed to complete its review. Enclosed is the NRG staff's request for additional information. The request for additional information was discussed with your staff on June 27, 2017, and it was agreed that your response would be provided within 30 days from the date of this letter.

Sincerely,

~~

Michael L. Marshall, Jr., Senior Pr9ject Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-220 and 50-410 Enclosure Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST NMP-RR-001 NINE MILE POINT NUCLEAR STATION, LLC EXELON GENERATION COMPANY, LLC NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-220 AND 50-41 O By letter dated March 7, 2017 (Agencywide Documents Access and Management System Accession No. ML17067A056), Exelon Generation Company, LLC (the licensee) submitted Relief Request NMP-RR-001 for the Nine Mile Point Nuclear Station, Units 1 and 2. The proposed relief request would authorize an alternative to performing 100 percent examination of the reactor pressure vessel nozzles listed in the request. The proposed alternative is to examine a minimum of 25 percent of the nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1."

The licensee stated in Attachment 3 of its submittal for Relief Request NMP-RR-001 that for extended operation to 60 years, the additional thermal cycle counts should be addressed for continued applicability of inspection relief request per ASME Code Case N-702. BWRVIP-108 and BWRVIP-241 are referenced by the licensee as the technical basis for its use of Code Case N-702. The licensee further explained that with respect to the Nine Mile Point, Unit 1, N2 nozzle (i.e., recirculation inlet nozzle), thermal fatigue crack growth due to extended operation to 60 years is not a controlling factor.

The U.S. Nuclear Regulatory Commission (NRG) staff notes that fatigue crack growth and the number of thermal transients was an input into the probabilistic fracture mechanics analyses performed in support of BWRVI P-108 and BWRVI P-241. Thus, it is not clear how the licensee determined that thermal fatigue crack growth and the additional thermal transients associated with extended operation to 60 years is not a controlling factor for the probabilistic fracture mechanics analysis for the Nine Mile Point, Unit 1, N2 nozzle.

The NRG staff has determined that the following additional information is required to complete its review:

• Provide a justification that demonstrates that fatigue crack growth and the additional thermal trainsets due to extended operation to 60 years for Nine Mile Point, Unit 1, is not a controlling factor and does not impact the probability of failure determined in BWRVIP-241 for the N2 nozzle.

Enclosure

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST TO UTILIZE ASME CODE CASE N-702 (CAC NOS. MF9381 AND MF9382) DATED JUNE 27, 2017 DISTRIBUTION:

Public RidsNrrDorlLpl1 RidsRgn1 MailCenter RidsACRS_MailCTR RidsNrrLALRonewicz RidsNrrPMNineMilePoint RidsNrrDeEvib Resource OYee, NRR ADAMS Access1on N um ber: ML17172A084 *b1y e-ma1*1 OFFICE DORL/LPL 1/PM DORL/LPL 1/LA DE/EVIB/BC* DORL/LPL 1/BC DORL/LPL 1/PM NAME MMarshall LRonewicz DRuland JDanna MMarshall DATE 06127 /2017 06/21/2017 06/02/2017 06/27/2017 06/27/2017 OFFICIAL RECORD COPY