LIC-15-0054, Response to NRC Request for Additional Information Regarding Aging Management Program for Reactor Vessel Intervals

Response to NRC Request for Additional Information Regarding Aging Management Program for Reactor Vessel Intervals

ML15103A642
Person / Time
Site:	Fort Calhoun
Issue date:	04/13/2015
From:	Cortopassi L Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-15-0054, TAC MF3412
Download: ML15103A642 (4)

Text

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UPPU Omaha Public Power District 444 South 1ftn Street Mall Omaha, NE 68102-2247 LIC-15-0054 April 13, 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington , DC 20555-0001 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285

Subject:

OPPD Response to NRC Request for Additional Information Regarding Aging Management Program For Reactor Vessel Internals (TAC No. MF3412)

References:

1. Letter from OPPD (L. P. Cortopassi) to NRC (Document Control Desk) , "Submittal of Reactor Vessel Internal (RVI) Component Aging Management Program (AMP) for Fort Calhoun Station (FCS), Unit No.1," dated September 27, 2012 (LIC-12-0144) (ML12276A005)

2. Letter from NRC (C. F. Lyon) to OPPD (L. P. Cortopassi) , "Fort Calhoun Station, Unit No.1 - Request for Additional Information RE:

Aging Management Program for Reactor Vessel Internals (TAC No.

MF3412)," dated March 3, 2015 (NRC-15-010) (ML15057A015)

This letter is submitted in response to an NRC request for additional information (Reference 2) regarding Omaha Public Power District's (OPPD) proposed aging management program for the Fort Calhoun Station, Unit No. 1, reactor vessel internals (Reference 1).

OPPD's response to the NRC RAI is attached.

This letter contains no regulatory commitments.

Employment with Equal Opportunity

U. S. Nuclear Regulatory Commission LI C-15-0054 Page2 If you should have any questions regarding this submittal or require additional information, please contact Mr. Bill R. Hansher, Principal Regulatory Engineer, at 402-533-6894.

Respectfully, ~

LoLJssl Site Vice President and CNO LPC/KGM/mle

Attachment:

OPPD Response to NRC Request for Additional Information Regarding Aging Management Program For Reactor Vessel Internals (TAC No. MF3412) c: M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S. M. Schneider, NRC Senior Resident Inspector

LIC-15-0054 Attachment Page 1 REQUEST FOR ADDITIONAL INFORMATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT N0.1 DOCKET NO. 50-285 In a letter dated September 27, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12276A005), Omaha Public Power District (OPPD, the licensee) submitted an aging management program (AMP) for the reactor vessel internals (RVls) at Fort Calhoun Station, Unit No. 1 (FCS). The Electric Power Research lnstitute's Materials Reliability Program report MRP-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," December 2008 (ADAMS Accession No. ML090160205), and its supporting reports were used as technical bases for developing FCS's AMP. The U.S. Nuclear Regulatory Commission (NRC) staff's safety evaluation (SE) for MRP-227-A was issued on December 16, 2011 (ADAMS Accession No. ML11308A770), with seven topical report conditions and eight applicant/licensee action items Section 6 2 of the licensee's September 27, 2012, submittal addressed the staff's eight action items. The following request for additional information (RAI) is related to the staff's Action Item 3 addressed in the staff's SE for the MRP-227-A.

RAl-2-1:

Section 6.2.3 of the licensee's September 27, 2012 submittal, included the following plant-specific reports which are related to the NRC staff's Action Item 3. The staff requested that the licensee provide a brief summary of the AMP related to the following RVI components. The summary should include: (a) the aging degradation; (b) the licensee's inspection methods of identifying aging effects; (c) the inspection results; and, (d) the frequency of subsequent inspections of these components during the extended period of operation.

• Plant-specific thermal shield integrity program - WCAP-17471-P, "OPPD Thermal Shield Integrity Program Task 2 Report," November 2011

• Plant-specific in-core instrumentation (ICI) thimble tube program - DAR-ME-05-15, Revision 00, "Recommendation for Replacement ICI Thimble Measurement,"

September 30, 2005 OPPD Response:

Thermal Shield The Thermal Shield is monitored for the aging mechanisms of cumulative fatigue damage, cracking, reduction of fracture toughness, changes in dimension, fatigue, and loss of preload .

Monitoring of neutron noise data is performed every 12 weeks (except during a refueling outage). A follow-up visual inspection (VT-3) to a 1992 repair will be performed in refueling outage (RFO) 28 scheduled for the fall of 2016 to confirm there is no further loss of preload.

Neutron noise data has been monitored since the plant restarted in 2013 and no issues have been identified. The frequency of future visual inspections will be determined after RFO 28.

LIC-15-0054 Attachment Page2 In-Core Instrumentation (ICI) Thimble Tubes The ICI thimble tubes are monitored for changes in dimensions, cracking, loss of material. and reduction in fracture toughness. The ICI tubes will be visually inspected (VT-3) in RFO 28, which will be the first time this inspection occurs as it is part of the inspections required by MRP-227-A. The maximum radial peaking factor is also monitored after every RFO to ensure it is within the limits described in the Core Operating Limits Report (COLA). The VT-3 inspection will determine which ICls need to be replaced. ICls are replaced within a 9 year interval (22 are scheduled for replacement in RFO 27 beginning in April 2015, and 6 are scheduled for replacement in RFO 28).