ML15057A015

Request for Additional Information, Materials Reliability Program (MRP)-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, Implementation Review

ML15057A015
Person / Time
Site:	Fort Calhoun
Issue date:	03/03/2015
From:	Lyon C Plant Licensing Branch IV
To:	Cortopassi L Omaha Public Power District
Lyon C
References
TAC MF3412
Download: ML15057A015 (3)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 3, 2015 Mr. Louis P. Cortopassi Site Vice President and Chief Nuclear Officer Omaha Public Power District Fort Calhoun Station 9610 Power Lane, Mail Stop FC-2-4 Omaha, NE 68008

SUBJECT:

FORT CALHOUN STATION, UNIT NO. 1 - REQUEST FOR ADDITIONAL INFORMATION RE: AGING MANAGEMENT PROGRAM FOR REACTOR VESSEL INTERNALS (TAC NO. MF3412)

Dear Mr. Cortopassi:

In a letter dated September 27, 2012, Omaha Public Power District, the licensee, submitted an aging management program (AMP) for the reactor vessel internals 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," 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, 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 NRC staff has reviewed the information provided in your application and determined that additional information is required in order to complete its formal review. The enclosed request for additional information is related to staff's Action Item 3 addressed in the staff's SE for the MRP-227-A. The enclosed questions were provided to 8. Hansher of your staff on February 27, 2015. Please provide a response to the enclosed questions within 45 days of the date of this letter. If you have any questions, please contact me at 301-415-2296 or via e-mail at Fred.Lyon@nrc.gov.

Sincerely, c.~

Carl F. Lyon, Project Manager Plant Licensing Branch IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. 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 requests 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) 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 Enclosure

ML15057A015 *memo dated **via email OFFICE NRR/DORL/LPL4-1/PM NRR/DORL/LPL4-1/LA NRR/DE/EVIB/BC* NRR/DORL/LPL4-1/BC(A) NRR/DORL/LPL4-1/PM NAME FLyon JBurkhardt** SRosenberg EOesterle FLyon DATE 2/27/15 2/26/15 2/24/15 3/2/15 3/3/15