ML101130223

Request for Additional Information for the Review of the Crystal River Unit 3 Nuclear Generating Plant, License Renewal Application

ML101130223
Person / Time
Site:	Crystal River
Issue date:	05/21/2010
From:	Kuntz R F License Renewal Projects Branch 2
To:	Franke J Florida Power Corp
Robert F. Kuntz, NRR/DLPM, 415-3733
References
TAC ME0274
Download: ML101130223 (8)

Text

May 21, 2010 Mr. Jon Franke, Vice President Crystal River Nuclear Plant (NA1B) ATTN: Supervisor, Licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, FL 34428-6708

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT, LICENSE RENEWAL APPLICATION (TAC NO. ME0274)

Dear Mr. Franke:

By letter dated December 16, 2009, Florida Power Corporation submitted an application pursuant to Title 10 of the Code of Federal Regulations Part 54, to renew the operating license for Crystal River Unit 3 Nuclear Generating Plant, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review. Further requests for additional information may be issued in the future. Items in the enclosure were discussed with Mr. Michael Heath, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-3733 or by e-mail at robert.kuntz@nrc.gov. Sincerely, /RA/

Robert F. Kuntz, Sr. Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosure:

As stated cc w/encl: See next page

ML101130223 OFFICE PM:RPB2:DLR LA:DLR BC:DLR:RARB BC:DRA:AFPB BC:RPB2:DLR PM:RPB2:DLR NAME RKuntz IKing GShukla AKlein DWrona RKuntz DATE 05/19/10 05/19/10 05/19/10 05/19/10 05/21/10 05/21/10 Crystal River Unit 3 Nuclear Generating Plant cc: Mr. R. Alexander Glenn Associate General Counsel (MAC-BT15A) Florida Power Corporation P.O. Box 14042 St. Petersburg, FL 33733-4042

Mr. James W. Holt Plant General Manager Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708 Mr. William A. Passetti, Chief Department of Health Bureau of Radiation Control 2020 Capital Circle, SE, Bin #C21 Tallahassee, FL 32399-1741

Attorney General Department of Legal Affairs The Capitol Tallahassee, FL 32304

Mr. Ruben D. Almaguer, Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive Tallahassee, FL 32399-2100

Chairman Board of County Commissioners Citrus County 110 North Apopka Avenue Inverness, FL 34450-4245 Mr. Stephen J. Cahill Engineering Manager Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708

Mr. Brian C. McCabe Manager, Nuclear Regulatory Affairs Progress Energy P.O. Box 1551 Raleigh, NC 27602-1551Mr. Daniel R. Westcott Supervisor, Licensing & Regulatory Programs Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428-6708

Senior Resident Inspector Crystal River Unit 3 U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River, FL 34428

Mr. Jack E. Huegel Manager, Nuclear Oversight Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708

Mr. David T. Conley Associate General Counsel II - Legal Dept. Progress Energy Service Company, LLC P.O. Box 1551 Raleigh, NC 27602-1551 Mr. Michael P. Heath Supervisor, License Renewal Progress Energy 8470 River Road, 2E Southport, NC 28461 Mr. Mark Rigsby Manager, Support Services - Nuclear Crystal River Nuclear Plant (SA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708 Mr. Robert J. Duncan II Vice President, Nuclear Operations Progress Energy P.O. Box 1551 Raleigh, NC 27602-155 REQUEST FOR ADDITIONAL INFORMATION LICENSE RENEWAL APPLICATION FOR CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302

RAI 2.4-1.1

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Background===

By letter dated August 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) staff requested additional information related to the Crystal River Unit 3 Nuclear Generating Plant (CR-3), license renewal application (LRA) including request for additional information (RAI) 2.4-1 which questioned the exclusion of fire barrier components from the scope of license renewal that appeared in a Safety Evaluation Report dated July 27, 1979. CR-3 letter dated September 30, 2009, responded to RAI 2.4-1 and stated that:

"As identified in Table 2.4.1-1, there are no fire doors, fire door penetration seals or interior fire hose stations in the Reactor Building. There are fire barrier assemblies which include Thermo-Lag fire barriers on conduits, junction boxes, transmitters, and penetrations encapsulated by stainless steel as discussed in response to RAI 2.3.3.36-3."

Issue:

National Fire Protection Association (NFPA) 14 "Standard for the Installation of Standpipe and Hose Systems," defines a Class III system as "A system that provides 1 1/2 in. (40 mm) hose stations to supply water for use by trained personnel and 2 1/2 in. (65mm) hose connections to supply a larger volume of water for use by fire departments." The CR-3 Final Safety Analysis Report (FSAR) Section 9.8.7.4, "Manual Fire Suppression Systems, "c. Standpipe and Hose Stations," states that, "The standpipes and hose station systems installed at CR-3 are Class II with the following exceptions: Reactor Building-Class III-" The response to RAI 2.4-1 seems to conflict the FSAR description and the NFPA 14 definition in terms of the presence of hose stations in the Reactor Building.

Request: Verify whether interior hose stations are present in the Reactor Building and if they are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an aging management review (AMR) in accordance with 10 CFR 54.21(a)(1). If they are excluded from the scope of license renewal and are not subject to an AMR provide justification for the exclusion. RAI 2.4-1.2

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Background===

CR-3 letter dated September 30, 2009, states on page 6 of 6 of Enclosure 3 that:

"As identified in Table 2.4.2-9, there are no fire barrier assemblies or interior fire hose stations in the Diesel Generator Building."

Issue: Section 5.4, "Diesel Generator Room," of the NRC Safety Evaluation Report, dated July 27, 1979, page 5-18 states that, "The diesel generator rooms and the control rooms are protected by a preaction automatic sprinkler system actuated by rate-compensating heat detector. In addition, the control room have smoke detectors that alarm in the plant control room. Portable fire extinguishers and interior hose are available for manual suppression."

Request: Verify whether interior hose mentioned in Section 5.4, "Diesel Generator Room," of the NRC Safety Evaluation Report, dated July 27, 1979, above are in the scope of license renewal in accordance with 10 CFR 54.4(a) and whether they are subject to an AMR in accordance with 10 CFR 54.21(a)(1). If they are excluded from the scope of license renewal and are not subject to an AMR provide justification for the exclusion.

RAI 4.3.3-5

Background

LRA Section 4.3.3 states that a bounding F en factor of 1.49 was used for the Alloy 600 component (incore instrumentation nozzle). It also states that the environmental effects for nickel alloys was obtained from Mehta, H. S., and S. R. Gosselin, "Environmental Factor Approach to Account for Water Effects in Pressure Vessel and Piping Fatigue Evaluations," Nuclear Engineering and Design, 1998. NUREG/CR-6335, which provides the statistical characterizations used to derive this F en factor for Alloy 600, states the fatigue S-N database for Alloy 600 is extremely limited and does not cover an adequate range of material and loading variables that might influence fatigue life. It further states that the data was obtained from relatively few heats of material and are inadequate to establish the effect of strain rate on fatigue life in air or of temperature in a water environment. NUREG/CR-6909 incorporates more recent fatigue data using a larger database for determining the F en factor of nickel alloys.

Issue: The reference used in the LRA to determine the environmental effects on nickel alloys may be non-conservative. The F en for nickel alloys based on NUREG/CR-6909 varies based on temperature, strain rate and dissolved oxygen. Based on actual plant operating conditions the F en factor can vary from a value 1.0 to 4.52 based on this methodology. Therefore, the cumulative usage factor (CUF) value for the incore instrumentation nozzle may be as high as 2.61 using the CUF presented in the LRA and the maximum F en derived from NUREG/CR-6909 which would exceed the design limit of 1.0 when considering environmental effects of reactor coolant during the period of extended operation.

Request: 1. Since the F en for nickel alloys can vary from 1.0 to 4.52 based on NUREG/CR-6909 and the CUF value may exceed the design limit of 1.0 for the incore instrumentation nozzle, justify using a value of 1.49 for the F en factor for this nickel alloy component.

2. Describe the current or future planned actions to update the CUF calculation with F en factor for the Alloy 600 component only, consistent with the methodology in NUREG/CR-6909. If there are no current or future planned actions to update the CUF calculation with F en factor for the Alloy 600 component consistent with the methodology in NUREG/CR-6909, provide a justification for not performing the update.

RAI B.2.23-1.1

Background

By letter dated December 1, 2009, the staff issued RAI B.2.23-1 requesting confirmation that the enhancements proposed for the External Surfaces Monitoring Program will specifically include physical manipulation and other investigative methods designed specifically to detect hardening and loss of strength in elastomers. By letter dated December 30, 2009, the applicant responded to the RAI. In that response, the applicant stated that the External Surfaces Monitoring Program will be used to visually inspect external surfaces and that the Internal Inspection of Miscellaneous Piping and Ducts Program will be used to visually and mechanically inspect internal surfaces.

Issue:

The staff notes that in polymeric materials the aging effect hardening/loss of strength can, depending on the environment to which exposed, initiate on either the internal or external surface of the component. Given that some components may be thick or rigid, it is not clear to the staff how mechanically inspecting a component from the interior surface alone will detect hardening/loss of strength which may initiate on the external surface. Additionally, unless the external surfaces monitoring program contains some requirements for the manual manipulation of polymeric materials, the staff is unsure how the applicant is specifically providing for the manual inspection of polymeric materials which are inspected only from the outside (such as those listed in tables 3.3.2-22, 3.3.2-23, 3.3.2-24, 3.3.2-25, 3.3.2-29, 3.3.2-48, 3.3.2-51, 3.3.2-52, and 3.3.2-54). Request: 1. Describe how the Internal Inspection of Miscellaneous Piping and Ducts Program will be used to detect hardening/loss of strength which originates on the external surfaces of polymeric materials or propose an alternate program which will accomplish this purpose.

2. Describe how the manual inspection of polymeric materials which are inspected only from the outside is addressed.

Letter to Jon Franke from Robert F. Kuntz dated May 21, 2010

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT, LICENSE RENEWAL APPLICATION (TAC NO. ME0274)

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