ML093010490
| ML093010490 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  (DPR-072) |
| Issue date: | 11/03/2009 |
| From: | Farideh Saba Plant Licensing Branch II |
| To: | Franke J Progress Energy Florida |
| Saba, F E, NRR/DORL/LPL2-2 301-415-1447 | |
| References | |
| BAW-2374, TAC ME1799 | |
| Download: ML093010490 (3) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 3, 2009 Mr. Jon A. Franke, Vice President Crystal River Nuclear Plant (NA2C)
ATTN: Supervisor, Licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, FL 34428-6708
SUBJECT:
CRYSTAL RIVER UNIT 3 - ONCE-THROUGH STEAM GENERATOR TUBE LOADS UNDER CONDITIONS RESULTING FROM POSTULATED BREAKS IN REACTOR COOLANT SYSTEM UPPER HOT-LEG LARGE-BORE PIPING (TAC NO. ME1799)
Dear Mr. Franke:
On June 25,2009, a public meeting was held between the U.S. Nuclear Regulatory Commission (NRC) staff and representatives of the Pressurized Water Reactor Owners Group (PWROG) at NRC Headquarters, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, regarding once-through steam generator tube loads under conditions resulting from postulated breaks in reactor coolant system upper hot-leg large-bore piping.
In the meeting, the NRC requested that each Babcock and Wilcox licensee submit a letter summarizing the information discussed at the meeting. The NRC staff followed up this verbal request with a letter dated July 31,2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092120002).
By letter dated August 25, 2009 (ADAMS Accession No. ML092390118), Florida Power Corporation (the licensee) provided its response for Crystal River, Unit 3. The NRC staff has reviewed the licensee's response and provides the following comments:
The staff has noted your response on the reporting requirements of Section 50.46, "Acceptance criteria for emergency core cooling systems for Iight water nuclear power reactors," of Title 10 of the Code of Federal Regulations.
We will contact you if we wish to have further discussions regarding your response.
Based on its review of this letter, the staff understands that the steam generators you are currently installing have been designed for large break loss-of-coolant accident (LBLOCA) loading conditions and that steam generator tube integrity will be maintained for all LBLOCAs (including those in the candy-cane region) as required by Technical Specification (TS) 3.4.16, "Steam Generator (OTSG) Tube Integrity," and TS 5.6.2.10, "Steam Generator (OTSG) Program."
The staff expects that any design/licensing basis documents, including the Updated Final Safety Analyses Report, will be updated, or already has been updated, to reflect your steam generator design and your management of steam
J. Franke
- 2 generator tube integrity for LBLOCA loads for the replacement steam generators.
This includes removing any reference to the approach discussed in the previously withdrawn PWROG topical report BAW-2374, Revision 2, "Risk Informed Assessment of Once-Through Steam Generator Tube Thermal Loads Due to Breaks in Reactor Coolant System Upper Hot Leg Large-Bore Piping," or similar approach.
This closes TAC No. ME1799. If you have any questions, please contact me at 301-415-1447 or bye-mail at Farideh.Saba@nrc.gov.
Sincerely, Farideh E. Saba, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket 1\\10. 50-302 cc: Distribution via ListServ
ML093010490 OFFICE NRRlLPL4/PM NRRlLPL2-21PM NRRlLPL4/LA NRRlDCI DRNAADB/BC NRRlDSSID NRRlLPL2-21BC NRRlLPL2-21PM NAME NKalyanam FSaba JBurkhardt KKarwoski RTaylor WRuland BySBahadur TBoyce FSaba DATE 10/29/09 11/02109 10/29/09 10/29/09 10/30/09 10/29/09 11/03/09 11/03/09