ML17206A101

Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)

ML17206A101
Person / Time
Site:	Nine Mile Point
Issue date:	07/24/2017
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF9381, CAC MF9382, NMP-RR-001
Download: ML17206A101 (5)

Text

Exelon July 24, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Units 1 and 2 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.55a Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-410

Subject:

Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)

References:

1) Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702)," dated March 7, 2017 2) Letter from M. Marshall (U.S. Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "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 In the Reference 1 letter, Exelon Generation Company, LLC (Exelon), requested relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components." Relief Request NMP-RR-001 proposed an alternative to the requirements contained in Table IWB-2500-1 concerning nozzle-to-vessel weld and nozzle inner radii examination requirements.

In the Reference 2 letter, the U.S. Nuclear Regulatory Commission requested additional information.

Attached is our response to that request. No regulatory commitments are contained in this letter. Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Sincerely , llflllR4 k;t::--James Barstow Director -Li c en s ing & Regulatory Affairs Exelon Generati on Company, LLC U.S. Nuclear Regulatory Commission Nine Mile Point Nuclear Station, Units 1 and 2 Submittal of Relief Request NMP-RR-001 Concerning Nozzle-to-Vessel Weld and Inner Radii Examinations (Use of Code Case N-702) July 24, 2017 Page2

Attachment:

Response to Request for Additional Information

-"NRC RAI on Nine Mile Point 1 Relief Request on Nozzle Inspection," Structural Integrity Associates, Inc., Report No. 1700909.401.RO, dated July 18, 2017 cc: Regional Administrator

-NRC Region I NRC Senior Resident Inspector

-NMP NRC Project Manager, NRR -NMP Attachment Response to Request for Additional Information

-"NRC RAI on Nine Mile Point 1 Relief Request on Nozzle Inspection," Structural Integrity Associates, Inc., Report No. 1700909.401.RO, dated July 18, 2017

©Structural Integrity Associates, Inc. July 18, 2017 Report No. 1700909.401.RO Quality Program: [gj Nuclear D Commercial 5215 Hellyer Ave. Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.strucbnt.com wwong@structinlcom

Subject:

NRC RAI on Nine Mile Point l Relief Request on Nozzle Inspection By letter dated March 7, 2017 (MLl 7067A056), Exelon submitted Relief Request NMP-RR-001 for the Nine Mile Point Nuclear Station, Units l and 2 based on SI Report No. 1501576.40 l, Revision l. The Nuclear Regulatory Commission (NRC) staff has determined that the following additional information is needed to complete its review of the request:

• Provide a justification that demonstrates that fatigue crack growth and the additional thermal transients due to extended operation to 60 years for Nine Mile Point, Unit l, is not a controlling factor and does not impact the Probability of Failure determined in BWRVIP-241 for the N2 nozzle. It was determined in SI Report No. 1501576.401 that fatigue crack growth is not a controlling factor for probability of failure for the Nine Mile Point Unit 1 N2 nozzle based on Reference 20 of the report, EPRI Letter 2012-138, dated August 31, 2012. In that letter, an assessment on the amount of fatigue crack growth was performed to assess the impact of additional thermal cycles due to license renewal operation.

A typical fracture toughness at the upper shelf, K1c, was taken as 200 ksi'1in. Using this upper shelf with a safety factor of '110, an allowable K1c was established.

Fatigue crack growth (FCG) was calculated using the allowable K1c as L\K and ASME Section XI, Appendix A crack growth laws. The resulting crack growth was calculated to be 0.24 inches for an assumed 300 cycles over 40 years. The stress corrosion cracking (SCC) over 40 years due to a constant maximum K equivalent to the allowable K1c was determined to be 33.88 inches. Thus, FCG is insignificant compared to SCC (several orders of magnitudes less) for typical nozzles in a cylinder where normal operating thermal expansion and pressure stresses are high. The Nine Mile Point 1 N2 nozzle was selected for analysis in BWRVIP-241 due to its bounding geometry.

However, it is located at the bottom head of the reactor, where pressure stresses are much lower than equivalent nozzles in the cylindrical portion of the vessel. This results in extremely low SCC and brings the SCC growth down to the same order of magnitude as FCG. As a confirmation that FCG is not a controlling factor for Nine Mile Point 1, a plant specific evaluation was performed where the BWRVIP-241 analyses on the Nine Mile Point 1 N2 nozzle was repeated with double the number of thermal fatigue cycles used by BWRVIP-241.

The result was no failures over l million simulations, which is the same result reported in BWRVIP-----------------

Toll-Free 877-474-7693

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Chicago, IL an -47 4-7693 Akron, OH 330*899-9753 Denver, CO 303-792-oon Auslln, TX 512-533-9191 San Diego, CA 858-455-6350 Charlotte, NC 704-597-5554 San Jose, CA 408-978-8200 Chattanooga, TN 423-553-1180 State College, PA 814-954-7776 Toronto, Canada 905-829-9817 Report No. 1700909.401.RO Julyl8,2017 Page 2 of2 241. This shows that additional thermal fatigue cycles from operation through 60 years on the Nine Mile Point Unit 1 N2 nozzle will not affect the resulting probability of failure. The Nine Mile Point l Nl nozzles are located in cylindrical portion of the vessel, so the conclusions based on the results of the analysis summarized above for EPRI Letter 2012-138 directly apply and fatigue is not controlling through 60 years. In addition, the Nl nozzles have fluence less than lxl0 17 n/cm2 through 60 years and the Nl nozzles meets the five criteria set forth by the NRC in the Safety Evaluation of BWRVIP-108 for inspection relief. Therefore, the results of B WRVIP-108 are considered applicable through 60 years. Prepared by:

7118/17 Wilson Wong Date Senior Engineer Reviewed by: -7/18/17 Approved by: Wilson Wong Senior Engineer 7/18/17 Date Verified by: 7/18/17 Date Associate Structural Integrity Associates, Inc.*