ML061380651

Technical Specification Bases (Tsb) Change

ML061380651
Person / Time
Site:	Oconee
Issue date:	05/10/2006
From:	Brandi Hamilton Duke Energy Carolinas, Duke Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML061380651 (56)

Text

BRUCE H HAMILTON Duke Vice President r Energy& Oconee Nuclear Station Duke Energy Corporation ONO1 VP / 7800 Rochester Highway Seneca, SC 29672 864 885 3487 May 10, 2006 864 885 4208 fax bhhamilton@duke-energy. corn U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk

Subject:

Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC Oconee Nuclear Station Docket Numbers 50-269, 270, and 287 Technical Specification Bases (TSB) Change Please see attached a revision to TSB 3.7.17, Spent Fuel Pool Ventilation System, to define the terminology 'recently irradiated fuel.' License Amendment 338, 339 and 339, Adoption of the alternate Source Term, added this definition to the affected bases. The bases of TS 3.7.17 was inadvertently missed. These changes are purely editorial.

Attachment 1 contains the new TSB pages, Attachment 2 contains the marked up version of the TSB pages and Attachment 3 contains NSD 228, Applicability Determination.

If any additional information is needed, please contact Reene Gambrell at 864-885-3364.

Very truly yours, B. H. Hamilton, Vice President Oconee Nuclear Site www. duke-energy. corn

U. S. Nuclear Regulatory Commission May 10, 2006 Page 2 cc: Mr. L. N. Olshan Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Mr. W. D. Travers, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mel Shannon Senior Resident Inspector Oconee Nuclear Station Mr. Henry Porter Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health & Environmental Control 2600 Bull Street Columbia, SC 29201

U. S. Nuclear Regulatory Commission May 10, 2006 Page 3 bcc: w/o attachments L. F. Vaughn C. J. Thomas - MNS R. D. Hart - CNS w/attachments Document Management ELL NSRB MR Coordinator (Ron Harris)

Licensing Working Group

May 9, 2006 RE: Oconee Nuclear Station Technical Specifications On April 11 2006, Station Management approved revisions to TSB 3.7.17, Spent Fuel Pool Ventilation System, to define the terminology 'recently irradiated fuel.' License Amendment 338, 339, and 339, Adoption of the Alternate Source Term, added this definition to the affected bases of TS. The bases of TS 3.7.17 was inadvertently missed. These changes are purely editorial.

Please revise your manuals as listed below.

Remove these pages Insert these pages

• • • THIS IS A CONTROLLED DOCUMENT, SUBJECT TO QA AUDIT. MANUALS ARE TO BE KEPT ACCURATE AND UPDATED AS SOON AS REVISIONS ARE RECEIVED.

• TSB LOEP Pages 1-17 TSB LOEP Pages 1-17 TSB Page B 3.7.17-1 TSB Page B 3.7.17-1 TSB Page B 3.7.17-2 TSB Page B 3.7.17-2 TSB Page B 3.7.17-3 TSB Page B 3.7.17-3 If you have any questions or problems, please call Reene Gambrell at 864-885-3364.

B. G. Davenport Regulatory Compliance Manager Regulatory Compliance By: Gail Joyner

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE LOEP1 BASES REVISION 4/12/06 LOEP2 BASES REVISION 12/14/04 LOEP3 BASES REVISION 12/14/04 LOEP4 338/339/339 6/1/04 LOEP5 338/339/339 6/1/04 LOEP6 338/339/339 6/1/04 LOEP7 336/336/337 11/5/03 LOEP8 BASES REVISION 7/25/05 LOEP9 BASES REVISION 12/20/05 LOEP10 BASES REVISION 3/31/05 LOEPi 1 BASES REVISION 1/17/06 LOEP12 BASES REVISION 1/17/06 LOEP13 BASES REVISION 4/12/06 LOEP14 338/339/339 6/1/04 LOEP15 338/339/339 6/1/04 LOEP16 338/339/339 11/23/05 LOEP17 BASES REVISION 11/23/05 i 320/320/320 9/26/01 ii 336/336/337 11/5/03 iii 330/330/331 4/30/03 iv 338/339/339 6/1/04 B 2.1.1-1 313/313/313 6/21/00 B 2.1.1-2 300/300/300 12/16/98 B 2.1.1-3 300/300/300 12/16/98 B 2.1.1-4 313/313/313 6/21/00 B 2.1.2-1 300/300/300 12/16/98 B 2.1.2-2 300/300/300 12/16/98 B 2.1.2-3 300/300/300 12/16/98 B 3.0-1 BASES REVISION 10/23/03 B 3.0-2 BASES REVISION 10/23/03 B 3.0-3 BASES REVISION 10/23/03 B 3.0-4 BASES REVISION 10/23/03 B 3.0-5 BASES REVISION 10/23/03 B 3.0-6 BASES REVISION 10/23/03 B 3.0-7 BASES REVISION 10/23/03 B 3.0-8 BASES REVISION 10/23/03 LOEPI

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.0-9 BASES REVISION 10/23/03 B 3.0-10 BASES REVISION 10/23/03 B 3.0-11 BASES REVISION 10/23/03 B 3.0-12 BASES REVISION 10/23/03 B 3.0-13 BASES REVISION 10/23/03 B 3.0-14 BASES REVISION 10/23/03 B 3.0-15 BASES REVISION 10/23/03 B 3.1.1-1 300/300/300 12/16/98 B 3.1.1-2 BASES REVISION 05/11/99 B 3.1.1-3 300/300/300 12/16/98 B3.1.1-4 300/300/300 12/16/98 B 3.1.2-1 300/300/300 12/16/98 B 3.1.2-2 300/300/300 12/16/98 B 3.1.2-3 300/300/300 12/16/98 B 3.1.2-4 300/300/300 12/16/98 B 3.1.2-5 300/300/300 12/16/98 B 3.1.3-1 BASES REVISION 06/02/99 B 3.1.3-2 BASES REVISION 03/27/99 B 3.1.3-3 300/300/300 12/16/98 B 3.1.3-4 300/300/300 12/16/98 B 3.1.4-1 BASES REVISION 12/14/04 B 3.1.4-2 BASES REVISION 12/14/04 B 3.1.4-3 BASES REVISION 12/14/04 B 3.1.4-4 BASES REVISION 12/14/04 B 3.1.4-5 BASES REVISION 12/14/04 B 3.1.4-6 BASES REVISION 12/14/04 B 3.1.4-7 BASES REVISION 12/14/04 B 3.1.4-8 BASES REVISION 12/14/04 B 3.1.4-9 BASES REVISION 12/14/04 B 3.1.5-1 300/300/300 12/16/98 B 3.1.5-2 300/300/300 12/16/98 B 3.1.5-3 300/300/300 12/16/98 B 3.1.5-4 ' 300/300/300 12/16/98 B 3.1.6-1 BASES REVISION 12/14/04 B 3.1.6-2 BASES REVISION 12/14/04 B 3.1.6-3 BASES REVISION 12/14/04 B 3.1.6-4 BASES REVISION 12/14/04 LOEP2

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.1.7-1 BASES REVISION 12/14/04 B 3.1.7-2 BASES REVISION 12/14/04 B 3.1.7-3 BASES REVISION 12/14/04 B 3.1.7-4 BASES REVISION 12/14/04 B 3.1.8-1 300/300/300 12/16/98 B 3.1.8-2 300/300/300 12/16/98 B 3.1.8-3 300/300/300 12/16/98 B 3.1.8-4 300/300/300 12/16/98 B 3.1.8-5 300/300/300 12/16/98 B 3.2.1-1 BASES REVISION 10/30/03 B 3.2.1-2 BASES REVISION 10/30/03 B 3.2.1-3 BASES REVISION 10/30/03 B 3.2.1-4 BASES REVISION 10/30/03 B 3.2.1-5 BASES REVISION 10/30/03 B 3.2.1-6 BASES REVISION 10/30/03 B 3.2.1-7 BASES REVISION 10/30/03 B 3.2.2-1 300/300/300 12/16/98 B 3.2.2-2 300/300/300 12/16/98 B 3.2.2-3 300/300/300 12/16/98 B 3.2.2-4 300/300/300 12/16/98 B 3.2.2-5 300/300/300 12/16/98 B 3.2.2-6 300/300/300 12/16/98 B 3.2.2-7 300/300/300 12/16/98 B 3.2.3-1 BASES REVISION 12/11/03 B 3.2.3-2 BASES REVISION 12/11/03 B 3.2.3-3 BASES REVISION 12/11/03 B 3.2.3-4 BASES REVISION 12/11/03 B 3.2.3-5 BASES REVISION 12/11/03 B 3.2.3-6 BASES REVISION 12/11/03 B 3.2.3-7 BASES REVISION 12/11/03 B 3.2.3-8 BASES REVISION 12/11/03 B 3.2.3-9 BASES REVISION 12/11/03 B 3.3.1-1 BASES REVISION 12/14/04 B 3.3.1-2 BASES REVISION 12/14/04 B 3.3.1-3 BASES REVISION 12/14/04 B 3.3.1-4 BASES REVISION 12/14/04 B 3.3.1-5 BASES REVISION 12/14/04 LOEP3

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.3.1-6 BASES REVISION 12/14/04 B 3.3.1-7 BASES REVISION 12/14/04 B 3.3.1-8 BASES REVISION 12/14/04 B 3.3.1-9 BASES REVISION 12/14/04 B 3.3.1-10 BASES REVISION 12/14/04 B 3.3.1-11 BASES REVISION 12/14/04 B 3.3.1-12 BASES REVISION 12/14/04 B 3.3.1-13 BASES REVISION 12/14/04 B 3.3.1-14 BASES REVISION 12/14/04 B 3.3.1-15 BASES REVISION 12/14/04 B 3.3.1-16 BASES REVISION 12/14/04 B 3.3.1-17 BASES REVISION 12/14/04 B 3.3.1-18 BASES REVISION 12/14/04 B 3.3.1-19 BASES REVISION 12/14/04 B 3.3.1-20 BASES REVISION 12/14/04 B 3.3.1-21 BASES REVISION 12/14/04 B 3.3.1-22 BASES REVISION 12/14/04 B 3.3.1-23 BASES REVISION 12/14/04 B 3.3.1-24 BASES REVISION 12/14/04 B 3.3.1-25 BASES REVISION 12/14/04 B 3.3.2-1 BASES REVISION 12/14/04 B 3.3.2-2 BASES REVISION 12/14/04 B 3.3.2-3 BASES REVISION 12/14/04 B 3.3.3-1 BASES REVISION 12/14/04 B 3.3.3-2 BASES REVISION 12/14/04 B 3.3.3-3 BASES REVISION 12/14/04 B 3.3.3-4 BASES REVISION 12/14/04 B 3.3.4-1 341/343/342 11/2/04 B 3.3.4-2 341/343/342 11/2/04 B 3.3.4-3 341/343/342 11/2/04 B 3.3.4-4 341/343/342 11/2/04 B 3.3.4-5 341/343/342 11/2/04 B 3.3.4-6 341/343/342 11/2/04 B 3.3.4-7 341/343/342 11/2/04 B 3.3.5-1 338/339/339 6/1/04 B 3.3.5-2 338/339/339 6/1/04 B 3.3.5-3 300/300/300 12/16/98 B 3.3.5-4 300/300/300 12/16/98 LOEP4

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.3.5-5 BASES REVISION 06/02/99 B 3.3.5-6 338/339/339 6/1/04 B 3.3.5-7 300/300/300 12/16/98 B 3.3.5-8 300/300/300 12/16/98 B 3.3.5-9 300/300/300 12/16/98 B 3.3.5-10 300/300/300 12/16/98 B 3.3.5-11 321/321/322 3/18/02 B 3.3.5-12 300/300/300 12/16/98 B 3.3.6-1 300/300/300 12/16/98 B 3.3.6-2 338/339/339 6/1/04 B 3.3.6-3 300/300/300 12/16/98 B 3.3.7-1 BASES REVISION 4/16/03 B 3.3.7-2 338/339/339 6/1/04 B 3.3.7-3 BASES REVISION 4/16/03 B 3.3.7-4 345/347/346 5/19/05 B 3.3.8-1 BASES REVISION 7/14/05 B 3.3.8-2 BASES REVISION 7/14/05 B 3.3.8-3 BASES REVISION 7/14/05 B 3.3.8-4 BASES REVISION 7/14/05 B 3.3.8-5 BASES REVISION 7/14/05 B 3.3.8-6 BASES REVISION 7/14/05 B 3.3.8-7 BASES REVISION 7/14/05 B 3.3.8-8 BASES REVISION 7/14/05 B 3.3.8-9 BASES REVISION 7/14/05 B 3.3.8-10 BASES REVISION 7/14/05 B 3.3.8-11 BASES REVISION 7/14/05 B 3.3.8-12 BASES REVISION 7/14/05 B 3.3.8-13 BASES REVISION 7/14/05 B 3.3.8-14 BASES REVISION 7/14/05 B 3.3.8-15 BASES REVISION 7/14/05 B 3.3.8-16 BASES REVISION 7/14/05 B 3.3.8-17 BASES REVISION 7/14/05 B 3.3.8-18 BASES REVISION 7/14/05 B 3.3.8-19 BASES REVISION 7/14/05 B 3.3.8-20 BASES REVISION 7/14/05 B 3.3.9-1 300/300/300 12/16/98 LOEP5

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.3.9-2 300/300/300 12/16/98 B 3.3.9-3 300/300/300 12/16/98 B 3.3.9-4 300/300/300 12/16/98 B 3.3.10-1 300/300/300 12/16/98 B 3.3.10-2 300/300/300 12/16/98 B 3.3.10-3 300/300/300 12/16/98 B 3.3.10-4 300/300/300 12/16/98 B 3.3.11-1 336/336/337 11/5/03 B 3.3.11-2 336/336/337 11/5/03 B 3.3.11-3 336/336/337 11/5/03 B 3.3.11-4 336/336/337 11/5/03 B 3.3.11-5 336/336/337 11/5/03 B 3.3.11-6 DELETE BASES REV 4/17/02 B 3.3.12-1 336/336/337 11/5/03 B 3.3.12-2 336/336/337 11/5/03 B 3.3.12-3 DELETE 320/320/320 9/26/01 B 3.3.13-1 336/336/337 11/5/03 B 3.3.13-2 336/336/337 11/5/03 B 3.3.13-3 336/336/337 11/5/03 B 3.3.13-4 336/336/337 11/5/03 B 3.3.14-1 300/300/300 12/16/98 B 3.3.14-2 300/300/300 12/16/98 B 3.3.14-3 300/300/300 12/16/98 B 3.3.14-4 300/300/300 12/16/98 B 3.3.15-1 BASES REVISION 3/11/04 B 3.3.15-2 BASES REVISION 3/11/04 B 3.3.15-3 BASES REVISION 3/11/04 B 3.3.16-1 338/339/339 6/1/04 B 3.3.16-2 338/339/339 6/1/04 B 3.3.16-3 338/339/339 6/1/04 B 3.3.16-4 300/300/300 12/16/98 B 3.3.17-1 300/300/300 12/16/98 B 3.3.17-2 BASES REVISION 03/27/99 B 3.3.17-3 300/300/300 12/16/98 B 3.3.18-1 300/300/300 12/16/98 B 3.3.18-2 300/300/300 12/16/98 B 3.3.18-3 300/300/300 12/16/98 B 3.3.18-4 300/300/300 12/16/98 LOEP6

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.3.19-1 BASES REVISION 7/12/01 B 3.3.19-2 BASES REVISION 7/12/01 B 3.3.19-3 BASES REVISION 7/12/01 B 3.3.19-4 BASES REVISION 7/12/01 B 3.3.20-1 300/300/300 12/16/98 B 3.3.20-2 300/300/300 12/16/98 B 3.3.20-3 300/300/300 12/16/98 B 3.3.20-4 300/300/300 12/16/98 B 3.3.21-1 300/300/300 12/16/98 B 3.3.21-2 300/300/300 12/16/98 B 3.3.21-3 300/300/300 12/16/98 B 3.3.22-1 BASES REVISION 03/27/99 B 3.3.22-2 300/300/300 12/16/98 B 3.3.23-1 300/300/300 12/16/98 B 3.3.23-2 300/300/300 12/16/98 B 3.3.23-3 300/300/300 12/16/98 B 3.3.23-4 300/300/300 12/16/98 B 3.3.24-1 320/320/320 9/26/01 B 3.3.25-1 336/336/337 11/5/03 B 3.3.25-2 Delete,336/336/337 11/5/03 B 3.3.25-3 Delete,336/336/337 11/5/03 B 3.3.25-4 Delete,336/336/337 11/5/03 B 3.3.25-5 Delete,336/336/337 11/5/03 B 3.3.25-6 Delete,336/336/337 11/5/03 B 3.3.26-1 336/336/337 11/5/03 B 3.3.26-2 Delete,336/336/337 11/5/03 B 3.3.26-3 Delete,336/336/337 11/5/03 B 3.3.27-1 336/336/337 11/5/03 B 3.3.27-2 Delete,336/336/337 11/5/03 B 3.3.27-3 Delete,336/336/337 11/5/03 B 3.3.28-1 BASES REVISION 3/19/02 B 3.3.28-2 BASES REVISION 3/19/02 B 3.3.28-3 BASES REVISION 3/19/02 B 3.3.28-4 BASES REVISION 3/19/02 B 3.4.1-1 313/313/313 6/21/00 B 3.4.1-2 309/309/309 1/18/00 B 3.4.1-3 300/300/300 12/16/98 B 3.4.1-4 309/309/309 1/18/00 B 3.4.1-5 300/300/300 12/16/98 B 3.4.2-1 300/300/300 12/16/98 B 3.4.2-2 300/300/300 12/16/98 B 3.4.3-1 4/17/01 LOEP7

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/A36 LIST OF EFFECTIVE PAGES PAGE AMENDMENT REVISION DATE B 3.4.3-2 BASES REVISION 4/17/01 B 3.4.3-3 BASES REVISION 4/17/01 B 3.4.3-4 BASES REVISION 4/17/01 B 3.4.3-5 BASES REVISION 4/17/01 B 3.4.3-6 BASES REVISION 4/17/01 B 3.4.3-7 BASES REVISION 4/17/01 B 3.4.3-8 BASES REVISION 4/17/01 B 3.4.4-1 300/300/300 12/16/98 B 3.4.4-2 300/300/300 12/16/98 B 3.4.4-3 300/300/300 12/16/98 B 3.4.4-4 300/300/300 12/16/98 B 3.4.5-1 BASES REVISION 3/25/04 B 3.4.5-2 BASES REVISION 3/25/04 B 3.4.5-3 BASES REVISION 3/25/04 B 3.4.5-4 BASES REVISION 3/25/04 B 3.4.6-1 300/300/300 12/16/98 B3.4.6-2 300/300/300 12/16/98 B 3.4.6-3 300/300/300 12/16/98 B 3.4.6-4 300/300/300 12/16/98 B 3.4.7-1 BASES REVISION 12/19/01 B 3.4.7-2 BASES REVISION 12/19/01 B 3.4.7-3 BASES REVISION 12/19/01 B 3.4.7-4 BASES REVISION 12/19/01 B 3.4.7-5 BASES REVISION 12/19/01 B 3.4.8-1 BASES REVISION 12/19/01 B 3.4.8-2 BASES REVISION 12/19/01 B 3.4.8-3 BASES REVISION 12/19/01 B 3.4.8-4 BASES REVISION 12/19/01 B 3.4.9-1 BASES REVISION 7/25/05 B 3.4.9-2 BASES REVISION 7/25/05 B 3.4.9-3 BASES REVISION 7/25/05 B 3.4.9-4 BASES REVISION 7/25/05 B 3.4.9-5 BASES REVISION 7/25/05 B 3.4.9-6 BASES REVISION 7/25/05 B 3.4.10-1 309/309/309 1/18/00 B 3.4.10-2 309/309/309 1/18/00 B 3.4.10-3 309/309/309 1/18/00 B 3.4.10-4 309/309/309 1/18/00 B 3.4.11-1 300/300/300 12/16/98 B 3.4.11-2 300/300/300 12/16/98 B 3.4.11-3 300/300/300 12/16/98 LOEP8

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT REVISION DATE B 3.4.11-4 300/300/300 12/16/98 B 3.4.11-5 300/300/300 12/16/98 B 3.4.12-1 BASES REVISION 12/20/05 B 3.4.12-2 BASES REVISION 12/20/05 B 3.4.12-3 BASES REVISION 12/20/05 B 3.4.12-4 BASES REVISION 12/20/05 B 3.4.12-5 BASES REVISION 12/20/05 B 3.4.12-6 BASES REVISION 12/20/05 B 3.4.12-7 BASES REVISION 12/20/05 B 3.4.12-8 BASES REVISION 12/20/05 B 3.4.12-9 BASES REVISION 12/20/05 B 3.4.12-10 BASES REVISION 12/20/05 B 3.4.12-11 BASES REVISION 12/20/05 B 3.4.12-12 BASES REVISION 12/20/05 B 3.4.13-1 300/300/300 12/16/98 B 3.4.13-2 BASES REVISION 05/11/99 B 3.4.13-3 300/300/300 12/16/98 B 3.4.13-4 300/300/300 12/16/98 B 3.4.13-5 300/300/300 12/16/98 B 3.4.14-1 335/335/336 9/29/03 B 3.4.14-2 340/342/341 9/2/04 B 3.4.14-3 340/342/341 9/2/04 B 3.4.14-4 335/335/336 9/29/03 B 3.4.14-5 335/335/336 9/29/03 B 3.4.14-6 335/335/336 9/29/03 B 3.4.15-1 300/300/300 12/16/98 B 3.4.15-2 300/300/300 12/16/98 B 3.4.15-3 300/300/300 12/16/98 B 3.4.15-4 300/300/300 12/16/98 B 3.4.15-5 300/300/300 12/16/98 B 3.5.1-1 BASES REVISION 1/11/05 B 3.5.1-2 BASES REVISION 1/11/05 B 3.5.1-3 BASES REVISION 1/11/05 B 3.5.1-4 BASES REVISION 1/11/05 B 3.5.1-5 BASES REVISION 1/11/05 B 3.5.1-6 BASES REVISION 1/11/05 B 3.5.1-7 BASES REVISION 1/11/05 B 3.5.1-8 BASES REVISION 1/11/05 B 3.5.2-1 BASES REVISION 3/31/05 B 3.5.2-2 BASES REVISION 3/31/05 LOEP9

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISEED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.5.2- 3 BASES REVISION 3/31/05 B 3.5.2- 4 BASES REVISION 3/31/05 B 3.5.2- 5 BASES REVISION 3/31/05 B 3.5.2- 6 BASES REVISION 3/31/05 B 3.5.2-'7 BASES REVISION 3/31/05 B 3.5.2-:8 BASES REVISION 3/31/05 B 3.5.2-!9 BASES REVISION 3/31/05 B 3.5.2-1 0 BASES REVISION 3/31/05 B 3.5.2-1 1 BASES REVISION 3/31/05 B 3.5.2-1 2 BASES REVISION 3/31/05 B 3.5.2-1 3 BASES REVISION 3/31/05 B 3.5.2-1 4 BASES REVISION 3/31/05 B 3.5.3- 1 340/342/341 9/2/04 B 3.5.3-l 340/342/341 9/2/04 B 3.5.3-2 340/342/341 9/2/04 B 3.5.3-' 340/342/341 9/2/04 B 3.5.3-! 5 340/342/341 9/2/04 B 3.5.3-( 340/342/341 9/2/04 B 3.5.3-4 340/342/341 9/2/04 B 3.5.3-4 340/342/341 9/2/04 B 3.5.3-S 340/342/341 9/2/04 B 3.5.3-1 0 Delete 9/2/04 B 3.5.4-1 323/323/324 4/22/02 B 3.5.4-2 300/300/300 12/16/98 B 3.5.4-3 323/323/324 4/22/02 B 3.5.4-4 323/323/324 4/22/02 B 3.5.4-5 323/323/324 4/22/02 B 3.5.4-6 323/323/324 4/22/02 B 3.6.1-1 300/300/300 12/16/98 B 3.6.1-2 300/300/300 12/16/98 B 3.6.1-3 300/300/300 12/16/98 B 3.6.1-4 BASES REVISION 1/25/01 B 3.6.1-5 Delete 12/16/98 B 3.6.2-1 BASES REVISION 01/31/00 B 3.6.2-2 300/300/300 12/16/98 B 3.6.2-3 300/300/300 12/16/98 B 3.6.2-4 300/300/300 12/16/98 B 3.6.2-5 300/300/300 12/16/98 B 3.6.2-6 BASES REVISION 03/27/99 B 3.6.2-7 BASES REVISION 03/27/99 LOEP10

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.6.3-1 BASES REVISION 12/18/01 B 3.6.3-2 BASES REVISION 12/18/01 B 3.6.3-3 BASES REVISION 12/18/01 B 3.6.3-4 BASES REVISION 12/18/01 B 3.6.3-5 BASES REVISION 12/18/01 B 3.6.3-6 BASES REVISION 12/18/01 B 3.6.3-7 BASES REVISION 12/18/01 B 3.6.3-8 BASES REVISION 12/18/01 B 3.6.3-9 BASES REVISION 12/18/01 B.3.6.3-10 BASES REVISION 12/18/01 B 3.6.4-1 300/300/300 12/16/98 B 3.6.4-2 300/300/300 12/16/98 B 3.6.4-3 BASES REVISION 03/27/99 B 3.6.5-1 BASES REVISION 12/14/04 B 3.6.5-2 BASES REVISION 12/14/04 B 3.6.5-3 BASES REVISION 12/14/04 B 3.6.5-4 BASES REVISION 12/14/04 B 3.6.5-5 BASES REVISION 12/14/04 B 3.6.5-6 BASES REVISION 12/14/04 B 3.6.5-7 BASES REVISION 12/14/04 B 3.6.5-8 BASES REVISION 12/14/04 B 3.6.5-9 BASES REVISION 12/14/04 B 3.6.5-10 BASES REVISION 12/14/04 B 3.7.1-1 BASES REVISION 1/11/05 B 3.7.1-2 BASES REVISION 1/11/05 B 3.7.1-3 BASES REVISION 1/11/05 B 3.7.2-1 BASES REVISION 10/13/03 B 3.7.2-2 BASES REVISION 10/13/03 B 3.7.2-3 BASES REVISION 10/13/03 B 3.7.2-4 BASES REVISION 10/13/03 B 3.7.2-5 BASES REVISION 10/13/03 B 3.7.3-1 BASES REVISION 1/17/06 B 3.7.3-2 BASES REVISION 1/17/06 B 3.7.3-3 BASES REVISION 1/17/06 B 3.7.3-4 BASES REVISION 1/17/06 B 3.7.4-1 BASES REVISION 1/17/06 B3.7.4-2 BASES REVISION 1/17/06 B3.7.4-3 BASES REVISION 1/17/06 B3.7.4-4 BASES REVISION 1/17/06 LOEPI 1

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.7.5-1 BASES REVISION 03/27/99 B 3.7.5-2 BASES REVISION 03/27/99 B 3.7.5-3 300/300/300 12/16/98 B 3.7.5-4 300/300/300 12/16/98 B 3.7.5-5 300/300/300 12/16/98 B 3.7.5-6 300/300/300 12/16/98 B 3.7.5-7 300/300/300 12/16/98 B 3.7.5-8 300/300/300 12/16/98 B 3.7.6-1 330/330/331 4/30/03 B 3.7.6-2 330/330/331 4/30/03 B 3.7.6-3 330/330/331 4/30/03 B 3.7.7-1 300/300/300 12/16/98 B 3.7.7-2 300/300/300 12/16/98 B 3.7.7-3 300/300/300 12/16/98 B 3.7.7-4 300/300/300 12/16/98 B 3.7.8-1 BASES REVISION 12/19/01 B 3.7.8-2 BASES REVISION 12/19/01 B 3.7.8-3 BASES REVISION 12/19/01 B 3.7.8-4 BASES REVISION 12/19/01 B 3.7.8-5 BASES REVISION 12/19/01 B 3.7.8-6 BASES REVISION 12/19/01 B 3.7.8-7 BASES REVISION 12/19/01 B 3.7.9-1 BASES REVISION 12/11/02 B 3.7.9-2 338/339/339 6/1/04 B 3.7.9-3 338/339/339 6/1/04 B 3.7.9-4 338/339/339 6/1/04 B 3.7.9-5 338/339/339 6/1/04 B 3.7.10-1 338/339/339 6/1/04 B 3.7.10-2 Deleted 338/339/339 6/1/04 B 3.7.10-3 Deleted 338/339/339 6/1/04 B 3.7.10-4 Deleted 338/339/339 6/1/04 B 3.7.11-1 BASES REVISION 1/17/06 B 3.7.11-2 BASES REVISION 1/17/06 B 3.7.11-3 BASES REVISION 1/17/06 B 3.7.12-1 323/323/324 4/22/02 B 3.7.12-2 323/323/324 4/22/02 LOEP12

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE 3.7.12-3 323/323/324 4/22/02 3.7.12-4 323/323/324 4/22/02 3.7.12-5 323/323/324 4/22/02 B 3.7.13-1 323/323/324 4/22/02 B 3.7.13-2 323/323/324 4/22/02 B 3.7.13-3 323/323/324 4/22/02 B 3.7.13-4 323/323/324 4/22/02 B 3.7.13-5 323/323/324 4/22/02 B 3.7.14-1 300/300/300 12/16/98 B 3.7.14-2 300/300/300 12/16/98 B 3.7.14-3 300/300/300 12/16/98 B 3.7.15-1 300/300/300 12/16/98 B 3.7.15-2 300/300/300 12/16/98 B 3.7.15-3 300/300/300 12/16/98 B 3.7.16-1 BASES REVISION 4/24/03 B 3.7.16-2 BASES REVISION 4/24/03 B 3.7.16-3 BASES REVISION 4/24/03 B 3.7.16-4 338/339/339 6/1/04 B 3.7.16-5 BASES REVISION 4/24/03 B 3.7.16-6 338/339/339 6/1/04 B 3.7.16-7 338/339/339 6/1/04 B 3.7.17-1 BASES REVISION 4/12/06 B 3.7.17-2 BASES REVISION 4/12/06 B 3.7.17-3 BASES REVISION 4/12/06 B 3.8.1-1 339/341/340 8/5/04 B 3.8.1-2 339/341/340 8/5/04 B 3.8.1-3 339/341/340 8/5/04 B 3.8.1-4 322/322/323 3/20/02 B 3.8.1-5 322/322/323 3/20/02 B 3.8.1-6 339/341/340 8/5/04 B 3.8.1-7 339/341/340 8/5/04 B 3.8.1-8 339/341/340 8/5/04 B 3.8.1-9 339/341/340 8/5/04 B 3.8.1-10 339/341/340 8/5/04 B 3.8.1-11 339/341/340 8/5/04 B 3.8.1-12 339/341/340 8/5/04 B 3.8.1-13 339/341/340 8/5/04 B 3.8.1-14 339/341/340 8/5/04 B 3.8.1-15 339/341/340 8/5/04 B 3.8.1-16 339/341/340 8/5/04 LOEP13

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.8.1-17 339/341/340 8/5/04 B 3.8.1-18 339/341/340 8/5/04 B 3.8.1-19 339/341/340 8/5/04 B 3.8.1-20 339/341/340 8/5/04 B 3.8.1-21 339/341/340 8/5/04 B 3.8.1-22 339/341/340 8/5/04 B 3.8.1-23 322/322/323 3/20/02 B 3.8.1-24 322/322/323 3/20/02 B 3.8.1-25 322/322/323 3/20/02 B 3.8.1-26 322/322/323 3/20/02 B 3.8.2-1 338/339/339 6/1/04 B 3.8.2-2 338/339/339 6/1/04 B 3.8.2-3 338/339/339 6/1/04 B 3.8.2-4 338/339/339 6/1/04 B 3.8.2-5 338/339/339 6/1/04 B 3.8.2-6 300/300/300 12/16/98 B 3.8.2-7 300/300/300 12/16/98 B 3.8.3-1 BASES REVISION 11/12/01 B 3.8.3-2 BASES REVISION 11/12/01 B 3.8.3-3 BASES REVISION 11/12/01 B 3.8.3-4 BASES REVISION 11/12/01 B 3.8.3-5 BASES REVISION 11/12/01 B 3.8.3-6 BASES REVISION 11/12/01 B 3.8.3-7 BASES REVISION 11/12/01 B 3.8.3-8 BASES REVISION 11/12/01 B 3.8.3-9 BASES REVISION 11/12/01 B 3.8.3-10 BASES REVISION 11/12/01 B 3.8.4-1 338/339/339 6/1/04 B 3.8.4-2 338/339/339 6/1/04 B 3.8.4-3 338/339/339 6/1/04 B 3.8.4-4 300/300/300 12/16/98 LOEP14

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.8.5-1 300/300/300 12/16/98 B 3.8.5-2 300/300/300 12/16/98 B 3.8.5-3 300/300/300 12/16/98 B 3.8.5-4 300/300/300 12/16/98 B 3.8.5-5 300/300/300 12/16/98 B 3.8.5-6 BASES REVISION 01/31/00 B 3.8.6-1 300/300/300 12/16/98 B 3.8.6-2 300/300/300 12/16/98 B 3.8.6-3 300/300/300 12/16/98 B 3.8.6-4 300/300/300 12/16/98 B 3.8.7-1 338/339/339 6/1/04 B 3.8.7-2 338/339/339 6/1/04 B 3.8.7-3 338/339/339 6/1/04 B 3.8.8-1 BASES REVISION 12/14/04 B 3.8.8-2 BASES REVISION 12/14/04 B 3.8.8-3 BASES REVISION 12/14/04 B 3.8.8-4 BASES REVISION 12/14/04 B 3.8.8-5 BASES REVISION 12/14/04 B 3.8.8-6 BASES REVISION 12/14/04 B 3.8.8-7 BASES REVISION 12/14/04 B 3.8.8-8 BASES REVISION 12/14/04 B 3.8.8-9 BASES REVISION 12/14/04 B 3.8.9-1 338/339/339 6/1/04 B 3.8.9-2 338/339/339 6/1/04 B 3.8.9-3 338/339/339 6/1/04 B 3.8.9-4 BASES REVISION 7/03/01 B 3.9.1-1 300/300/300 12/16/98 B 3.9.1-2 300/300/300 12/16/98 B 3.9.1-3 300/300/300 12/16/98 LOEP15

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.9.2-1 300/300/300 12/16/98 B 3.9.2-2 300/300/300 12/16/98 B 3.9.2-3 BASES REVISION 06/02/99 B 3.9.2-4 300/300/300 12/16/98 B 3.9.3-1 338/339/339 6/1/04 B 3.9.3-2 338/339/339 6/1/04 B 3.9.3-3 338/339/339 6/1/04 B 3.9.3-4 338/339/339 6/1/04 B 3.9.3-5 338/339/339 6/1/04 B 3.9.4-1 BASES REVISION 4/25/02 B 3.9.4-2 BASES REVISION 4/25/02 B 3.9.4-3 BASES REVISION 4/25/02 B 3.9.4-4 BASES REVISION 4/25/02 B 3.9.5-1 BASES REVISION 12/19/01 B 3.9.5-2 BASES REVISION 12/19/01 B 3.9.5-3 BASES REVISON 12/19/01 B 3.9.5-4 BASES REVISION 12/19/01 B 3.9.6-1 338/339/339 6/1/04 B 3.9.6-2 338/339/339 6/1/04 B 3.9.6-3 338/339/339 6/1/04 B3.9.7-1 309/309/309 1/18/00 B3.9.7-2 309/309/309 1/18/00 B3.9.7-3 309/309/309 1/18/00 B 3.10.1-1 BASES REVISION 11/23/05 B 3.10.1-2 BASES REVISION 11/23/05 B 3.10.1-3 BASES REVISION 11/23/05 B 3.10.1-4 BASES REVISION 11/23/05 B 3.10.1-5 BASES REVISION 11/23/05 B 3.10.1-6 BASES REVISION 11/23/05 B 3.10.1-7 BASES REVISION 11/23/05 B 3.10.1-8 BASES REVISION 11/23/05 B 3.10.1-9 BASES REVISION 11/23/05 B 3.10.1-10 BASES REVISION 11/23/05 B 3.10.1-11 BASES REVISION 11/23/05 B 3.10.1-12 BASES REVISION 11/23/05 B 3.10.1-13 BASES REVISION 11/23/05 B 3.10.1-14 BASES REVISION 11/23/05 B 3.10.1-15 BASES REVISION 11/23/05 B 3.10.1-16 BASES REVISION 11/23/05 B 3.10.1-17 BASES REVISION 11/23/05 LOEP16

OCONEE NUCLEAR STATION TECHNICAL SPECIFICATIONS-BASES REVISED 04/12/06 LIST OF EFFECTIVE PAGES PAGE AMENDMENT BASES REVISION DATE B 3.10.1-18 BASES REVISION 11/23/05 B 3.10.2-1 300/300/300 12/16/98 B 3.10.2-2 300/300/300 12/16/98 B 3.10.2-3 300/300/300 12/16/98 B 3.10.2-4 300/300/300 12/16/98 B 3.10.2-5 300/300/300 12/16/98 B 3.10.2-6 300/300/300 12/16/98 LOEP 17

Attachment #1 Proposed Bases revision Remove Page Insert Page B 3.7.17-1 B 3.7.17-1 B 3.7.17-2 B 3.7.17-2 B 3.7.17-3 B 3.7.17-3

SFPVS B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Pool Ventilation System (SFPVS)

BASES BACKGROUND Ventilation air for the Spent Fuel Pool Area is supplied by an air handling unit which consists of roughing filters, steam heating coil, cooling coil supplied by low pressure service water, and a centrifugal fan. In the normal mode of operation, the air from the Spent Fuel Pool Area is exhausted directly to the unit vents by the general Auxiliary Building exhaust fans. The filtered exhaust system consists of a single filter train and two 100 percent capacity vane axial fans. The filter train utilized is the Reactor Building Purge Filter Train. The Unit 2 Reactor Building purge filter train is used for the combined Unit 1 and 2 Spent Fuel Pool Ventilation System, The Unit 3 Reactor Building purge filter train is used for the Unit 3 SFP Ventilation System. The filter train is comprised of prefilters, HEPA filters, and charcoal filters. To control the direction of air flow, i.e., to direct the air from the Fuel Pool Area to the Reactor Building Purge Filter Train, a series of pneumatic motor operated dampers are provided along with a crossover duct from the Fuel Pool to the filter train.

The SFPVS is discussed in the UFSAR, Section 9.4.2, (Ref. 1).

APPLICABLE The analysis of the limiting fuel handling accident, the cask drop SAFETY ANALYSES accident, given in Reference 2, assumes that a certain number of fuel assemblies are damaged. The DBA analysis for the cask drop accident, does not assume operation of the SFPVS in order to meet the requirements of 10 CFR 50.67 (Ref. 4). These assumptions and the analysis are consistent with the guidance provided in Regulatory Guide 1.183 (Ref. 3).

The SFPVS does not satisfy the criteria in 10 CFR 50.36. The SFPVS is retained in this specification for ALARA purposes.

LCO With the adoption of the alternate source term and the installation of various plant modifications, SFPVS is not credited in dose analysis calculations. Therefore, there are no specific operability requirements for this system.

OCONEE UNITS 1, 2, & 3 B 3.7.17-1 BASES REVISION DATED 04/12/06 l

SFPVS B 3.7.17 BASES LCO An SFPVS train is considered OPERABLE when its associated:

(continued)

1. Fan is OPERABLE;

2. Filter trains are intact; and

3. Ductwork and dampers are OPERABLE, and air flow can be maintained.

APPLICABILITY During movement of recently irradiated fuel assemblies (i.e., fuel that has occupied part of a critical reactor core within the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) in the fuel handling area, the SFPVS shall be OPERABLE.

ACTIONS A.1 and A.2 With one SFPVS train inoperable, the OPERABLE SFPVS train must be started immediately with its discharge through the associated reactor building purge filter or recently irradiated fuel movement in the spent fuel pool suspended. This action ensures that the remaining train is OPERABLE, and that any active failures will be readily detected.

If the system is not placed in operation, this action requires suspension of recently irradiated fuel movement, which precludes a fuel handling accident. This action does not preclude the movement of recently irradiated fuel assemblies to a safe position.

OCONEE UNITS 1, 2, & 3 B 3.7.1 7-2 BASES REVISION DATED 04/12/06 l

SFPVS B 3.7.17 BASES ACTIONS B.1 (continued)

When two trains of the SFPVS are inoperable during movement of recently irradiated fuel in the spent fuel pool, the unit must be placed in a condition in which the LCO does not apply. This Action involves immediately suspending movement of recently irradiated fuel assemblies in the spent fuel pool. This does not preclude the movement of recently irradiated fuel to a safe position.

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not severe, testing each train within 31 days prior to moving recently irradiated fuel assemblies provides an adequate check on this system. The system is no longer credited in dose analysis calculations and is not required to maintain 10 CFR 50.67 dose limits, but is required for ALARA purposes.

SR 3.7.17.2 This SR verifies that the required SFPVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.

REFERENCES 1. UFSAR, Section 9.4.2.

2. UFSAR, Section 15.11.

3. Regulatory Guide 1.183.

4. 10 CFR 50.67.

5. Dose Calculations.

OCONEE UNITS 1, 2, & 3 B 3.7.17-3 BASES REVISION DATED 04/12/06 l

Attachment #2 Markup of current Bases

SFPVS B 3.7.17 BASES LCO An SFPVS train is considered OPERABLE when its associated:

(continued)

1. Fan is OPERABLE;

2. Filter trains are intact; and

3. Ductwork and dampers are OPERABLE, and air flow can be maintained.

APPLICABILITY uring ment (i.e., fuel that has }

occupied part of a critical reactor core within the previous 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) in the I fu din are SE VSsal PRBE ACTIONS A.1 and A.2 With one SFPVS train inoperable, the OPERABLE SFPVS train must be started immediately with its discharge through the associated reactor building purge filter or recently irradiated fuel movement in the spent fuel pool suspended. This action ensures that the remaining train is OPERABLE, and that any active failures will be readily detected.

If the system is not placed in operation, this action requires suspension of recently irradiated fuel movement, which precludes a fuel handling accident. This action does not preclude the movement of recently irradiated fuel assemblies to a safe position.

OCONEE UNITS 1, 2, & 3 B 3.7.17-2 ON3SES REVISION DATED XX/X

Attachment #3 NSD 228, Applicability Determination

APPENDIX C 228 APPLICABILITY DETERMINATION Rev. 2 DUKE POWER SITE itUNIT(S) i Oconee E McGuire El CataIba Y unit i UnitP2 Une o3 ACTIVITY TfTLEIDOCUMENTIREVISION su Implementation of UFSAR, TS bases, SLC, SLC bases, and DBD changes that support changes to SFPVS d pueto of alternate methodology. Changes approved by NRC (ref. Amendment 338 339, and 339, Issued June 1,2004). I vI PAR_ -E PROCESSD RE Will Implementation of the above activity require a change to the:

I. Technical Specifications (TS) or Operating License? Ij NO LI YES If yes, process as a license amendment per NSD 221.

2. Quality Assurance Topical? 0 NO LI YES If yes, seek assistance from Nuclear Regulatory industry Affairs in NAD.

3. Security Plans? i NO LI YES If yes, process per the Nuclear Security Manual.

4. Emergency Plan? 3 NO []YES If yes, process per the Emergency Planning Functional Area Manual.

5. Inservice Testing Program Plan? 0 NO L YES If yes, process per site IST Program for ASME code compliance and related facility changes.

6. Inservice Inspection Program Plan? i NO [ YES If yes, process per Materials, Metallurgy and Piping Inservice Inspection FAM for ASME code compliance and related facility or procedure changes.

7. Fire Protection Program Plan? iI NO L YES If yes, evaluate activity in accordance with NSD 320.

8. ISFSI Safety Analysis Report? iNO D YES If yes, refer to NSD 21.

9. Regulatory Commitments? 0 NO [ YES If yes, process per NSD 214..

10. Code of Federal Regulations? i NO [ YES If yes, contact the Regulatory Compliance group.

11. Programs and manuals listed in the Administrative Section It NO LI YES If yes, contact the Regulatory Compliance group.

of the TS that require review under a TS change process other than I0CFR50.59?

P=TARCRO5 CAkILf w

12. Does the activity involve a procedure, governed by NSD 703 that has been excluded from the 10 CFR 50.59 process 0 NO E YES per NSD 703, Appendix N and the exclusion status remains valid?

13. Does the activity involve an administrative procedure governed by NSD 100 that does not contain information RI NO L YES regarding the operation and control of Structures, Systems and Components?

14. Does the activity involve software and data governed by NSD 800 that NSD 800 excludes from the 10 CFR 50.59 0 NO D YES Process? Consult NSD 800 for assistance.

15. Does the activity involve a type of Engineering Change that NSD 301 excludes from the 10 CFR 50.59 Process? 0 NO I YES Consult NSD 301 for assistance.

16. Does the activity involve (a) maintenance activities that restore SSCs to their as-designed condition (including JO NO EL YES activities that implement approved design changes) or (b) temporary alterations supporting maintenance that will be in effect during at-power operations for 90 days or less?

17. Does the activity involve a UFSAR modification that NSD 220 excludes from the 10 CFR 50.59 Process. Consult WI NO n YES NSD 220 for assistance.

18. Does the activity involve NRC and/or Duke Power approved changes to the licensing basis? E NO 0 YES

19. Are ALL aspects of the activity bounded by one or more "YES" answers to questions I through 18, above? n NO W YES (Print Name) Dan Harrelson (Sign) Date: 11/15/2005 Applicability Determination Preparer

^0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTOK D.C. 2A 50 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 338 TO RENEWED FACILITY OPERATING LICENSE DPR-38 AMENDMENT NO. 339 TO RENEWED FACILITY OPERATING LICENSE DPR-47 AND AMENDMENT NO. 339 TO RENEWED FACILITY OPERATING LICENSE DPR-55 DUKE ENERGY CORPORATION OCONEE NUCLEAR STATION. UNITS 1. 2. AND 3 DOCKET NOS. 50-269. 50-270. AND 50-287

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC) dated October 16,2001, as supplemented by letters dated May 20, September 12, and November 21, 2002, September 22, and November 20, 2003, and February 18 and April 1.4, 2004, Duke Energy Corporation (the licensee) submitted a request for full implementation ol the alternative source term (AST) for changes to the Oconee Nuclear Station, Units 1,2, and 3, technical specifications (TSs). The requested change would incorporate revisions to the TSs resulting from the use of an AST that revises the radiological consequence analyses for two -design-basis accidents (DBAs) based on the AST, the loss-of-coolant accident (LOCA), and the fuel-handling accident (FHA). The supplements dated May 20, September 12, and November 21, 2002, and February 18 and April 14, 2004, provided clarifying information that did not change the scope of the October 16, 2001, submittal.

Inthe October 16,2001, submittal, the licensee requested, among other things, to remove the operational requirements of the penetration room ventilation system (PRVS) and spent fuel pool ventilation system (SFPVS) from the TSs. The proposed no significant hazards consideration (NSHC) was published Inthe FederalRegister(FR) on January 22, 2002 (67 FR 2922). The supplement dated September 22.2003, changed the scope of the October 16, 2001, submittal based on discussions with the NRC staff. The revised proposed NSHC was published in the FR on October 14, 2003 (68 FR 59215). Inthe supplement dated September 22, 2003, the licensee requested to (1)retain the PRVS and SFPVS in the TSs without crediting the removal of fission products by these systems in the radiological consequence analyses for the site boundaries and for the control room, and (2)adopt. echnical Specification Task Force (TSTF)

Change Traveler TSTF-51, Revision 2.

On November 20, 2003, based on further discussions with the NRC staff, the licensee resubmitted the request to remove the PRVS from the TSs and to adopt TSTF-51 for the

SFPVS with the addition of one surveillance requirement change concerning operational testing of the SFPVS trains. The revised proposed NSHC was published In the FR on December 9, 2003 (68 FR 68660).

Specific TS changes requested are listed in Section 4.0, "Technical Specification Changes," of this Safety Evaluation (SE).

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulations, design criteria, and guides to evaluate the licensee's request The current radiological consequence analyses for the DBA for Oconee are based on the TD-14844 accident source term. In this license amendment request, the licensee requested a full-scope Implementation of the AST, as described in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,* and pursuant to Title 10 of the Code of FederalRegulations (10 CFR) Section 50.67, "Accident Source Term." The use of an AST is addressed in 10 CFR 50.67, which provides a mechanism for licensed power reactors to replace the traditional source term used Intheir DBA radiological consequence analyses.

The NRC staff evaluated the radiological consequences of affected DBAs against the dose criteria specified in 10 CFR 50.34(b)(2); these criteria are 0.25 Sieverts (Sv), or 25 rem, total effective dose equivalent (TEDE) at the exclusion ared boundary (EAB) for any 2-hour period following the onset of the postulated fission product release, 0.25 Sv (25 rem) TEDE at the outer boundary of the low population zone (LPZ) for the duration of exposure to the release cloud, and 0.05 Sv (5 rem) TEDE In the control room.

The NRC staff used applicable guidance in Standard Review Plan (SRP) Section 15.0.1, "Radiological Consequence Analyses Using Altemative Source Term," and RG 1.183. Other relevant regulatory documents used by the NRC staff are (1)General Design Criterion (GDC) 19, "Control Room," of Appendix A to 10 CFR Part 50 and NUREG-0737, Item III.D.3.4, as they relate to maintaining the control room Ina safe, habitable condition under accident conditions by providing adequate protection against radiation and toxic gases, and (2) 10 CFR 50.36(c)(2), 1Umiting CQnditions for Operation" in TSs, as it relates to criteria for Including an item as a limiting condition for operation In TSs.

Oconee Updated Final Safety Analysis Report (UFSAR) Section 3.1.17, Criterion 17, "Monitoring Radioactivity Releases (Category B):' provides the licensee's interpretation and intent of agreement with the original Atomic Energy Commission GDC proposed in 1967 for monitoring radioactivity releases that the licensee adopted as part of the Oconee initial plant design-basis.

Oconee UFSAR Section 3.1.70, Criterion 70, "Control of Releases of Radioactivity to the Environment (Category B)," states that the facility design shall include those means necessary to maintain control over the plant's radioactive effluents, whether gaseous, liquid, or solid.

Appropriate holdup capacity shall be provided for retention of gaseous, liquid, or solid effluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design

for radioactivity control shall be justified: a) on the basis of 10 CFR Part 20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur, and b) on the basis of 10 CFR Part 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence, except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents.

RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Ucensing Basis," provides a definition of defense-in-depth and guidance on using a qualitative approach to risk assessment.

Defense-in-depth is a long standing and widely accepted design philosophy used in the NRC regulatory process. Defense-in-depth Is used to assure that there are multiple barriers or systems available to assure that the public and operating staffs are protected from the release of radioactive elements during both accident and normal operating events. Defense-in-depth is Invoked In RG 1.183, which provides the guidance on the use of the AST.

TSTF-51 is an Industry-initiated and NRC-accepted method for allowing some engineered ssafety feature (ESF) systems and components to be non-operable during a part of the refueling outage, subject to a defined decay period and acceptable shutdown administrative controls.

3.0 TECHNICAL EVALUATION

3.1 Loss-of-Coolant Accident The current radiological consequence analysis for the postulated LOCA is based on the accident source term described In TiD-I 4844 and It is provided in UFSAR Section 15.15, "Maximum Hypothetical Accident." To demonstrate that the ESFs designed to mitigate the radiological consequences at Oconee will remain adequate after implementing the changes requested in this license amendment, the licensee re-analyzed the offsite and control room radiological consequences of the postulated LOCA. The licensee has implemented an AST in.

this reanalysis.

The licensee submitted the results of Its offsite and control room dose calculations and provided the major assumptions and parameters used In its dose calculations. As documented in Its submittals, the licensee has determined that after Implementation of the changes requested in this license amendment with use of an AST, the existing ESF systems at Oconee will still provide assurance that the radiological consequences of the postulated LOCA at the EAB, in the LPZ, and in the control room will meet the acceptable radiation dose criteria specified in 10 CFR 50.67(b)(2). As part of the implementation of the AST, the TEDE acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole-body and thyroid dose guidelines of 10 CFR 100.11 and GDC 19.

The NRC staff has reviewed the licensee's analyses and has performed an independent confirmatory radiological consequence dose calculation for the following three potential fission product release pathways:

(1) reactor building (RB) leakage, (2) leakage from ESF systems outside containment, and

(3) emergency core cooling system (ECCS) back-leakage to the borated water storage tank (BWST) through check valves In the normal suction line from the BWST.

3.1.1 Reactor Building Leakage The Oconee RB completely encloses the reactor coolant system. It Is designed for an Internal pressure of 59 psig and the current RB design-basis leak rate Is 0.25 percent by weight per day

(%perday). Forthe radiological consequence analysis, this rate Is followed by 0.125% perday after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA for the duration of the accident (30 days). In addition, 50% of the 0.250%per day leakage Is processed by the PRVS. The remaining 50% of the 0.25% per day leakage Is assumed to be released directly to the environment.

The licensee proposed to lower the RB leak rate specified In the current TS Section 5.5.2 to 0.20'1iper day from 0.25% per day. For the radiological consequence analysis, the licensee also proposed to lower the RB leak rate to 0.1% per day from 0.1 25% per day after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA for the duration of the accident.

In Its original submittal dated October 16,2001, the licensee proposed to delete the PRVS from the TSs. The PRVS is an ESF system designed to collect and process RB leakage to minimize the release of radioactivity to the environment following a LOCA. This system uses high efficiency particulate air (HEPA) filters and charcoal adsorbers.

In a subsequent supplement dated September 22, 2003, the licensee revised its original request to retain the PRVS In the TS without crediting:the removal of fission products by this system in the LOCA radiological consequence analyses for the site boundaries and for the control room. In the supplement dated November 20, 2003, the licensee proposed again to delete the PRVS from the TS without crediting the removal of fission products.

As a result of not crediting the removal of fission products by the PRVS, the licensee proposed to delete the requirement from the TSs to measure the RB leakage rate to the penetration room. The licensee assumed all leakage (100 percent) from the RB Is released directly to the environment without filtration by the PRVS.-

The fission products in the RB atmosphere following the postulated LOCA are mitigated by natural deposition of fission products In aerosol form and by the RB spray system (RBSS). The licensee stated in its letter dated September 12, 2002, that it has conservatively neglected aerosol deposition in the RB and excluded any credit for the removal of fission products in aerosol form by natural deposition In the radiological consequence re-analyses. The radiological consequence analyses performed by the licensee showed that Oconee would still meet the relevant dose criteria specified In 10 CFR 50.67 without any credit for removing fission products by natural deposition processes in the RB.

The RBSS is an ESF system and is designed to provide RB cooling and fission product removal Inthe RB following the postulated LOCA. The RBSS consists of two spray pumps that are automatically started by a high RB pressure signal (15 psig). The licensee assumed that spray flow is Initiated within 96 seconds from the initiation of the postulated LOCA. During the RBSS operation, the licensee assumed a mixing rate of two unsprayed volumes per hour between the sprayed and unsprayed portions of the RB atmosphere that is consistent with the guidance provided in RG 1.183. This mixing rate Is 30500 cubic feet per minute.

I:

Two pumps start taking suction initially from the BWST and initiate building spray through two spray headers until the water in the BWST reaches a pre-set low level 25 minutes after the accident. The spray pump suction Is then transferred manually to the RB sump and the spray water Is re-circulated until the spray operation Isterminated 112.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident.

Two pump operation is more conservative than one pump operation Inthe radiological analysis since two pump operation will draw down the water Inthe BWST earlier, starting recirculation, which provides a lower Iodine removal efficiency. The licensee used the models and guidance provided InRG 1.183 to determine the removal rates by the RBSS for Iodine In elemental form and fission products in particulate form. Therefore, the NRC staff finds these removable rates acceptable. The major parameters and assumptions used by the licensee, including the spray removal rates, are listed inTable 2.

3.1 2-Post-LOCA Leakage From Engineered Safety Features Outside Containment Any leakage water from ESF components located outside the RB releases fission products during the recirculating phase of long-term core cooling following a postulated LOCA. The leakage from ESF components occurs Inthe auxiliary building. The licensee assumed the leakage into the auxiliary building to start at 25 minutes following a postulated LOCA when the RBSS switches Its water suction from the BWST to the RB sump. The leakage rate is assumed to be 50 gallons per hour, which is 2 times the allowable leakage limit. The licensee assumed that 10 percent of the total iodine activity Inthe leaked fluid becomes airborne and Is released directly to the environment throughout the entire period of the accident. These assumptions (ESFieak rate and iodine release fraction) are consistent with the guidelines provided in RG 1.183, and therefore, the NRC staff finds they are acceptable. The major parameters and assumptions used by the licensee are listed InTable 2.

3.1.3 Emergency Core Cooling System Back-Leakage to Borated Water Storage Tank During a postulated LOCA, the suction water source for the ECCS is switched from the BWST to the RB sump. The licensee assumed the leakage into the auxiliary building to start at 25 minutes following a postulated LOCA when the ECCS switches its water suction from the BWST to the RB sump. In this configuration, check valves in the normal suction line from the BWST provide Isolation between the RB sump and the BWST. The licensee assumed, In its radiological consequence analysis, 5 gallons per minute (gpm) back-flow leakage through check valves into the BWST from the RB sump. The licensee stated in its letter dated May 20, 2002, that the back-leakage to the BWST is tested every reactor outage and that the 5 gpm leakage value assumed in the radiological consequence analyses bounds actual test results.

In determining the radiological dose calculation through this pathway, the licensee developed its own methodology which is partly based on a model described InNUREG/CR-5950, "Iodine Evolution and pH Control.* The licensee calculated and submitted time dependent BWST water temperature, iodine concentrations in the BWST liquid and vapor, and the BWST water pH.

Using these calculated values, the licensee determined Iodine release rates from the BWST atmosphere to the environment. The NRC staff reviewed the model used and calculation performed by the licensee for fission product release through this pathway. The NRC staff finds the licensee's release model and assumptions to be conservative, and therefore, acceptable.

The major parameters and assumptions used by the licensee are listed in Table 2.

3.1.4 Radiological Consequence of Loss-of-Coolant Accident The licensee re-evaluated the radiological consequences resulting from the postulated LOCA using the AST and concluded that the radiological consequences at the EAB and LPZ are within the dose criteria specified In 10 CFR 50.67. To verify the licensee's radiological consequence analyses, the NRC staff performed its confirmatory radiological consequence dose calculation and found Its results are also within the dose criteria specified In 10 CFR 50.67. Although the NRC staff performed Its Independent radiological consequence dose calculation as a means of confirming the licensee's results, the NRC staff's acceptance Is based on the licensee's analyses. The results of the licensee's radiological consequence calculation are provided InTable I and the major parameters and assumptions used by the licensee and found acceptable by the NRC staff are listed In Tables 2 and 4. The radiological consequences of the LOCA at the EAB and at the LPZ calculated by the licensee and by the NRC staff are within the dose criteria specified in 10 CFR 50.67 and the control room dose limit as established by GDC 19.

32 Fuel-Handling Accident The current radiological consequence analysis for the postulated design-basis FHA is based on the accident source term described In TID-1 4844, and It Is provided In UFSAR Section 15.11.

In the TS change request, the licensee proposed to revise TS Section 3.9.3 to allow the containment air lock doors, the equipment hatch and each penetration providing direct access from the containment atmosphere to the outside atmqsphere to remain open during fuel movement and refueling operations. The licensee reevaluated the radiological consequences resulting from a postulated FHA In the spent fuel pool (SFP) area and in the containment with the personnel air locks open, the equipment access hatch open, and the containment penetrations open providing direct access from the containment atmosphere to the environment. The licensee Implemented the AST in its re-evaluation of a FHA. Each of the' Oconee unit's containment Is equipped with two containment air lock doors and an equipment access hatch.

The licensee considered twoJfuel-handling ev for 16tsthe postulated FHA: (1) the drop of a single fuel assembly in the containment or SIP. and (2)the drop of a fuel transport cask or an Independent spent fuel storage Installation (ISFSI) transfer cask (with multiple fuel assemblies)

In the SFP. The licensee concluded that the radiological consequences resulting from the postulated FHA In the SFP area and in the containment with open personnel air locks, the equipment access hatch open, and the containment penetrations providing direct access from the containment atmosphere to the environment are within the dose acceptance criteria specified in SRP 15.0.1 for the EAB and In 10 CFR 50.67 for the control room.

The licensee reached this conclusion as a result of:

(1) Implementing the AST, (2). taking no credit for removal of fission products by the PRVS, SFPVS, or the RB purge exhaust filtration system as proposed in the TS change request, (3) assuming no containment isolation (open air lock doors, equipment hatch, and containment penetrations) as proposed in this TS change request,

(4) using an overall decontamination factor of 183 for Iodine in elemental and organic forms In the SFP water with minimum water depth of 21.34 feet consistent with the guidelines provided in RG 1.183, (5) using the ARCON96 model documented In RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," In determining the control room intake air dispersion factors (X/Q values) (see Section 3.4 below),

(6) taking credit for dual control room air intake locations, reducing its X/Q values by assuming that wind flow is In the direction of the intake with the poorer atmospheric dispersion conditions and that this intake Is drawing In 60 percent of the air (see Section 3.4 below),

(7) using new relocated control room air Intake locations on the northeast and southeast comers of the turbine building roof as proposed Inthis TS change request, (8) assuming all fuel rods in one fuel assembly with an axial power peaking factor of 1.65 and a peak rod average fuel bumup of 62,000 megawatt days per metric ton uranium (MWDIMTU) are damaged to the extent that its entire gap activity Inventory of the damaged fuel rods is released to the surrounding water for single fuel assembly drop

-event consistent with the guidelines provided in-RG 1.183 (the corresponding maximum linear heat generation rate is 6.0 kW per foot based on 418.6 effective full power days, 3 cycles, an active fuel height of 11.86 feet, and 208 pins per fuel assembly), and (9) using a fission product decay period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (time period from the reactor shutdown to the first fuel movement) for single fuel assembly drop event.

For multiple fuel assembly events involving a drop of a fuel transport cask or an ISFSI transfer cask Inthe SFP, the number of fuel assemblies involved in these events varies depending on the location of the cask drop (i.e., the Unit 1 and 2, or Unit 3, SFP) and the type of cask dropped (i.e., transport cask or ISFSI cask). Oconee is designed with two SFPs: the Unit 1 and 2 SFP and the Unit 3 SFP. In its letter dated May 30.2000, the licensee provided a table summarizing the number of fuel assemblies that would be involved in each cask drop event in each SFP ranging from 518 to 1024 fuel assemblies. Since the number of fuel assemblies involved In these events is greater than the amount of fuel recently discharged from a core, (i.e., 177 fuel assemblies per core), the licensee assumed two different decay times for each event one for the fuel recently discharged from the core (55 to 70 days) and one for the other fuel involved in the event (one year). The NRC staff finds the licensee's assumption conservative and, therefore, acceptable.

The licensee performed 12 different radiological consequence cases for these FHA events, using a combination of different type of cask dropped (i.e., transport cask or ISFSI cask),

different location of the cask drop (i.e., the Unit 1 and 2, or Unit 3, SFP), and the different control room air intakes (i.e., Unit 1 and 2, or Unit 3, control room). Oconee has two control rooms, one for Unit 1and 2 and one for Unit 3. The licensee determined that (1) the bounding single fuel assembly FHA event is the FHA in either the Unit I and 2, or Unit 3, SFP to the Unit

1 and 2 control room air Intake, and (2)the bounding multiple-fuel assembly FHA event is the transport cask drop event Ineither the Unit 1 and 2, or Unit 3, SFP to the Unit 1 and 2 control room air Intake.

The NRC staff reviewed the methods, parameters, and assumptions that the licensee used In Its radiolgical dose consequence analyses. To verify the licensee's radiological consequence assessments, the NRC staff performed confirmatory radiological consequence dose calculations for the FHA events that will produce the greatest radiological consequence. The NRC staff finds, and agrees with the licensee, that the bounding single fuel assembly FHA event Isthe FHA in.the Unit 1 and 2 SFP to the Unit 1 and 2 control room air intake, and (2)the bounding multiple fuel assembly FHA event Is the transport cask drop event Inthe Unit 1 and 2 SFP to the Unit 1 and 2 control room air Intake. The radiological consequences calculated by the NRC staff are well within the dose criterion specified in 10 CFR 50.67 (5 rem TEDE In the control room) and meet the dose acceptance criterion specified Inthe SRP 15.0.1 (a 6.3 rem TEDE at the EAB).

Even though the NRC staff performed Its confirmatory dose calculations, the NRC staff's acceptance Isbased on the licensee's analyses. The results of the licensee's radiological consequence calculations are provided InTable 1 and the major parameters and assumptions used by the-licensee, which are acceptable to the NRC staff, are listed inTable 3. The radiological consequences at the EAB, at the LPZ, and in the control room calculated by the licensee are all well within the dose criterion specified in 10 CFR 50.67 and meet the dose acceptance criterion specified in GDC 19.

3.3 Control Room Habitability The Oconee control rooms are located Inthe auxiliary building. Units 1 and 2 have a shared control room while Unit 3 has a separate control room. The control room habitability system Includes two control room ventilation system booster fan trains (CRVSBFrs). The CRVSBFT is provided to maintain a positive control room pressure to ensure outward leakage, preventing unfiltered air inleakage into the control room. The CRVSBFT consists of, among other things, two separate outside air booster fans, HEPA filters and charcoal adsorbers. Each booster fan is designed to provide 1350 CFM. The licensee assumed that the control room operator will start the control room ventilation system (CRVS) booster fans within 30 minutes of the DBA.

The licensee requested a revision to the acceptance criteria for the CRVSBFT charcoal adsorbers inTS Section 5.5.12 to greater than 97.5 percent radioactive methyl Iodide removal from the current requirement of greater than 90 percent.

The licensee committed to relocate the existing control room outside air Intakes for the shared Unit 1and 2 control room and for the separate Unit 3 control room. The control room outside air intakes will be moved from the roof of the auxiliary building to the northeast and southeast corners of the turbine building roof. Dual intakes will be installed for each control room. The Oconee station presently has one outside Intake for each control room. Moving the intakes further from the source-term release points would reduce the control room operator dose based on the ARCON96 dispersion model (see Section 3.4 below). Installation of dual Intakes, as described inthe licensee's planned modification, would divide the dose between the two control rooms, assuming that one intake for each control room drew in contaminated air and the second intake for each control room drew in clean air. The licensee's radiological

consequence analyses to support this TS change request used the relocated control room outside air Intakes and dual Intake design to meet the relevant dose acceptance criterion.

On June 12,2003, the NRC staff Issued Generic Letter (GL) 2003-01, 'Control Room Habitability.0 This GL Identifies NRC staff concerns regarding the reliability of current surveillance testing to identify and quantify control room Inleakage, and requests licensees to confirm the most Irmiting unfiltered Inleakage into their control room envelope. On December 9, 2003, the licensee submitted a *60-dayr response to this GL for Oconee. In this submittal, the licensee stated that tracer gas tests were performed at Oconee In 1998 and 2001 for both the Unit 1 and 2 control room, and the Unit 3 control room, to confirm that the inleakage Into the control room envelope Is less than or equal to the value assumed in the Oconee radiological consequence analysis. The purpose of the 2001 re-test was to demonstrate the effectiveness of the CRVS sealing program.

The test methodology and analyses in determining the unfiltered air Inleakage rate to the control room used In the 1998 and 2001 tests were based on the American Society for Testing and Materials Standard E741-95, Standard Test Method for Determining Air Change Rate In a Single Zone by Means of a Tracer Gas Dilution.* The licensee provided the test results obtained In 1998 In Its letter dated October 16, 2001, and provided the test results obtained in 2001 in Its letter dated May 20,2002. The test results from the 2001 test were not available when the October 16, 2001, letter was submitted.

Both test results are provided in Table 5. The 1998 test resulted in 80 cfm for the Unit 1 and Unit 2 control room and 73 cfm for the Unit 3 control room. The uncertainty associated with the testing and analyses was 555 cfm for the Unit 1 and Unit 2 control room and + 25 cfm for the Unit 3 control room. The 2001 test resulted in zero measured Inleakage (Ocfm) to both control rooms; the uncertainty associated with the testing and analyses Is + 18 cfm for the Unit 1 and Unit 2control room and + 13 cfm for the Unit 3 control room.

In its radiological consequence analyses, the licensee has chosen a value of 40 cfm as the current design-basis control room unfiltered air inleakage rate based on the most recent (performed in 200-1) -tracer gas test result (in spite of zero measured inleakage to both control rooms) in conjunction with the total allowable ECCS leakage rate of 25 gph. The licensee stated that the selection of bounding values for the control room unfiltered inleakage assumed In the radiological analyses provides the licensee with margin to accommodate changes in input assumptions that could be required to account for possible future plant operational changes (e.g., increase in ECCS leakage flow, Imbalances ventilation system air flow, reductions in filtration efficiencies).

The licensee provided a sensitivity study with a table showing various combinations of the control room unfiltered air inleakage rates and the total allowable ECCS leakage rates meeting the dose acceptance criteria. The licensee indicated that, based on the result of the next tracer gas test planned In the Control Room Habitability Program, the licensee may change to a different combination of the control room unfiltered air Inleakage rates and the total allowable ECCS leakage rates, still meeting the dose acceptance criteria for the EAB, LPZ, and control room. The NRC staff finds the licensee's approach to be acceptable. The NRC staff also used the 40 cim control room unfiltered air inleakage rate and the 25 gph total allowable ECCS leakage rate in its confirmatory radiological consequence analyses.

For the first 30 minutes Into the postulated LOCA prior to the operation d the CRVS In the emergency pressurization mode In its radiological consequence analyses, the licensee has chosen to use the bounding unfiltered air Inleakage rate obtained from the 1998 test (1150 cSm), with the CRVS In the normal operational mode for the Unit 1 and 2 control room, since Itis more conservative than that obtained In the 2001 test. The NRC staff also used this leakage value In its confirmatory radiological consequence analyses.

The licensee re-evaluated the control room habitabilitywith the application of the AST using the bounding 40cfm unfiltered airinleakage Into the control room and concluded that the radiological consequences to the control room operator resulting from the postulated LOCA and FHA were within the 5 rem TEDE criterion specified In 10 CFR 50.67. To verify the licensee's radiological consequence assessments, the NRC staff performed Its confirmatory radiological consequence dose calculations for the control room operator and finds its results were also within the 5 rem TEDE criterion specified In10 CFR 50.67. Although the NRC staff performed Its Independent radiological consequence dose calculation as a means of confirming the licensee's results, the NRC staff's acceptance Is based on the licensee's analyses.

The results of the licensee's radiological consequence calculation are providedin Table1 and the major parameters and assumptions used by the licensee and acceptable to the NRC staff are listed in Tables 2 and 4. The radiological consequences for the control room operator calculated by the licensee and by the NRC staff are all within the dose criteria specifiedIn 10 CFR 50.67.

Although the NRC staff has reviewed the licensee's response to GL 2003-01 for Oconee, as well as those received from other licensees, follow-on regulatory action has not been decided at this time. Nonetheless, the NRC staff has determined that thereIs reasonable assurance that the Oconee control rooms will be habitable during a LOCA a'nd FHA and that these amendments may be approved prior to the NRC staff's formal review of the licensee's response to the GL for Oconee. The NRC staff bases this determination on (1) the Oconee tracer gas test results, which confirmed that the inleakage into the control room envelopeis less than the value assumed Inthe Oconee radiological consequence analyses; (2)the NRC staff's confirmIng analysis;:and (3) the resulting control room operator doses, which meet the acceptable dose limit The NRC staff's approval of these amendments does not relieve the licensee of addressing the Information requests In GL 2003-01 for Oconee and does not Imply that the NRC staff would necessarily find the analysis In these amendments acceptable as a response to Information request 1(a)In GL 2003-01.

3.4 Atmospheric Dispersion Factors 3.4.1 Meteorological Data The licensee used five years of on-site meteorological data collected during calendar years 1991 through 1995 to estimate the atmospheric relative concentration (X/Q)values used in the radiological consequence analyses. These data were measured at 10 and 60 meters above grade at the Oconee station. The licensee stated that the program is maintained to comply with RG 1.23, 'Onsite Meteorological Programs," and has further described the Oconee meteorological measurement program as follows. The tower in use since 1988 is located in a clearing, west of the plant. The area around the tower has been kept free of obstructions to

ensure good exposure of the Instruments. Weekly system checks and semi-annual calibration checks are performed to ensure that the instruments are maintained within specifications.

During the semi-annual calibrations, all instruments on the tower are replaced with newly certified Instruments. Data were reviewed by data reviewers, and validated and edited by the licensee's In-house Certified Consulting Meteorologist prior to archival. Annual data recoveries were approximately 96 percent during the first three years and about 93 percent during the second two years of the 1991 through 1995 period.

The NRC staff performed a review of the data, using the methodology described in NUREG 0917, "Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data," on meteorological data quality assurance. Further review was performed using a computer spreadsheet. Examination of the data revealed some occurrence of wind data -remaining unchanged for two or more consecutive hours. While the occurrence seemed somewhat higher to the NRG staff than expected by the meteorological factors Identified by the licensee, even with the uncertainty, the NRC staff estimates that the recoveryis still, at a minimum, near 90 percent and the data recovery uncertainties should not have a significant impact on the licensee's X/Q estimates. Although there were several occurrences of data outages of more than a week's duration, most outages were of relatively short duration.

With respect to atmospheric stability, the A stability class wasinfrequently reported to occur for a relatively long period of time, in excess of a dayin several cases. Class A stability conditions were also reported to occurinfrequently at night The licensee attributed this to factors such as cold air advection aloft or post-frontal air mass changes. Regardless of the cause, the NRC staff judges the reported occurrences to have aninsignificant effect on the X/Q estimates for the dose. assessment described above; Wind speed and direction frequency occurrence at each of the two levels, were fairly similar from year to year, with 1991 tending toward slightly lower wind speeds at both levels. Lower levelwinds were predominately from the northeast and southwest with secondary flow from the northwest. Upper level winds were most often from the north through east northeasterly sectors and from the southwest.

3.4.2 EAB and LPZ Relative Concentration Estimates With respect to the EAB and LPZ X/Q values, the licensee has used values approved by the Directorate of Licensing, U.S. Atomic Energy Commission, previously in the "Safety Evaluation of the Oconee Nuclear Power Station, Units 2 and 3,issued in 1973. Because Unit1 isin dose proximity to Unit 2 when compared with the distances to the EAB and LPZ, these values should be acceptable for application to Unit 1. The NRC staff did not directly review the X/Q values as a part of this evaluation, but notes that the calculation methodology used by the licensee at that time, about 30 years ago, was different than current guidance and NRC staff practice. However, the NRC staff performed a comparison calculation using the 1991 through 1995 meteorological data and the PAVAN methodology that Is based upon RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants." The results of the NRC staff calculations support the numeric acceptability of the licensee's EAB and LPZ X/Qestimates for this amendment. The licensee's values are listed in Table 4.

3.4.3 Control Room Relative Concentration Estimates The licensee used the ARCON96 methodology discussed in RG 1.194 to make calculations for all possible source-receptor pairs and selected the most limiting cases for use In the dose assessment. All postulated releases were considered as ground level point releases except those from the equipment hatch and fuel handling building roll-up door. The licensee considered the postulated hatch and door releases as ground level diffuse releases. The NRC staff did not make a judgement as to the acceptability of the diffuse release assumptions, but Bid make comparison calculations assuming the releases were ground level point releases.

Resultant values were slightly higher than assuming a diffuse source, but were still bounded by the X/Q values from releases from the limiting plant vent used Inthe radiological consequence analyes.

Inits initial assessment, the licensee divided the limiting X/Q values by a factor of two. Use of this reduction factor Is not part of the current Oconee licensing basis. The licensee based the reduction on plans to modify the CRVS to use dual Intakes, rather than a single airintake for each control room asIs the current design. This assumed that both of the dualintake channels would perform continuously and simultaneously during the course of an accident toindividually provide one-half of the needed air and that air from at least one of the twointakes could be assumed to be uncontaminated. However, the licensee acknowledged that the final design of the control room intakes and airIntake flow rates may vary slightly from the assumptions used in the preliminary design assessment. Therefore, thelicensee performed additional calculations using projected design variations that wotld potentially impact the X/O values.

Thisincluded an assumption that theintake locations would be10 feet higher than Intake locations used In the preliminary design and anIntake flow rate Imbalance of 55 and 45 percent, rather than each intake providing 50 percent of the flow.

The intake flow imbalance assumption was based on operational experience with dual intakes at one of the licensee's other plants. The licensee also assumed that wind flow Is in the direction of theintake with the poorer atmospheric dispersion conditions and that this intakeis drawing in 55 percent of the flow. In its letter dated September 22, 2003, the licensee provided revised control room X/Qvalues. The resultant values for postulated releases from the plant vent and BWST used in the limiting dose calculations for these proposed amendments are listed in Table 4. Present NRC guidance permits reduction of the X/Q values for dual intakes when the intakes for eachindividual control room are adequately separated to provide a low contamination intake and when these intakes are adequately designed. The NRC staff finds the physical separation of the intakes for each Individual control room adequate. The NRC staff also finds the X/Q values listed in Table 4 acceptable predicated on the assumption that both of the dual intake channels for each individual control room are designed and operated to be capable of performing continuously and simultaneously to individually provide the air flow assumed in the licensee's analysis and that air from at least one of the two intakes could be assumed to be uncontaminated for the duration of an accident.

4.0 TECHNICAL SPECIFICATION CHANGES The NRC staffhas reviewed these proposed changes to the TSs and has made the following assessments based on the descriptions of the changes provided by the licensee, the licensee's UFSAR, and the application of the regulations identified in Section 2.0.

t, 4.1 TS 3.3.6 Engineered Safeguards Protective System (ESPS) Manual Initiation Umifing Conditions for Operations (LCO) 3.3.6 c Isbeing changed to delete the reference to "Penetration Room Ventilation."

The licensee states that the PRVS will not be credited for control room and off-site doses based on the revised radiological analyses of the Maximum Hypothetical Accident (MHA). The NRC staff has confirmed that the PRVS does not need to be credited Inthe MHA analysis and that the analysis results are acceptable. As such, the PFIVS Is not on the primary success path In the mitigation of a DBA. Thus, the ESPS manual Initiation channels for the PRVS specified In LCO 3.3.6 are not required to be OPERABLE to support the mitigation of a DBA. The NRC staff has found the proposed change to remove reference to the PRVS to be acceptable.

4.2 TS 3.7.10 Penetration Room Ventilation System (PRVS)

The licensee proposed to remove this section from the TS. The PRVS will not be credited for evaluating potential control room and off-site doses. This change results in an operational efficiency that Is achievable from Implementing the AST. The revised radiological analyses of the MHA are performed without taking credit for the PRVS filters and the results of this analysis show that the off-site and control room doses remain below the guidance provided InRG 1.183.

Removwal otphis system from the TSs eliminates the requirement to demonstrate the effectiveness of this system Inoperation. This simplifies testing design and performance tasks.

The NRC staff notes that the licensee is requesting that the PRVS filter system be removed from the TS. As currently licensed, the assumed unfiltered leakage from the containment is 0.1375% per day (based on 90% filter efficiency, and 0.25% per day leakage Inwhich 50% of the leakage passes through the PRVS filters). With the proposed TS changes, the assumed unfiltered leakage per day Increases to 0.2% per day. This is a 45 percent increase in the assumed unfiltered containment leakage that is released to the environment. However, the resulting dose-calculations show that the release will remain within the licensing regulatory limits with substantial margin.

The NRC staff determined that. if the licensee takes credit for higher control room filter efficiency and for its newly installed dual control room intakes, the impact of a 45 percent larger unfiltered release from the containment can be mitigated for the control room dose and the licensee's AST analysis would produce acceptable dose results without taking credit for the PRVS filters. The licensee in the May 20, 2002, submittal stated that 50 percent of the unfiltered leakage was assumed to be released from the plant vent. This statement Implies that the fans must be operational to drive the unfiltered containment leakage to the plant vent.

The plant vent that was used as the release point Inthe MHA. A release from this location is considered a ground release and that its dispersion factor Isconservative with respect to the equipment hatch release dispersion factor. Thus, a vent release bounds other potential release paths (i.e., the equipment hatch release path) and provides additional assurance of acceptability.

The NRC staff evaluated the removal of the PRVS filters to determine the Impact on defense-in-depth. The penetration rooms serve as a secondary containment for containment leakage. The filters provide a barrier to the release of radiation to the environment from

containment penetration leakage during an accident. The NRC staff concludes that the Impact on defense-in-depth Isacceptable based on the low.frequency of the Initiating event and the Impact on doses resulting from the release, which are well below the regulatory limits.

UFSAR Section 3.1.17, Criterion 17, requires that releases of radioactive elements from the facility be monitored. Use of the PRVS provides for monitoring a portion of the containment releases during an accident by routing the flow to the plant vent. The licensee conservatively takes credit for the release from the plant vent InIts MHA analysis. This provides reasonable assurance that this part of the release will be monitored.

UFSAR Section 3.1.70 commits to those means necessary to maintain control over the plant radioactive effluents. it states specifically that the design for radioactivity control will be based on the-basis of 10 CFR Part 100 dosage level guidelines for potential reactor accidents of exceeding low probability. The MHA Inwhich the containment could be pressurized Is an accident of low probability. The RG 1.183 limits referenced by the licensee are taken from 10 CFR 60.67 and are the equivalent of 10 CFR Part 100 dosage limits for application of the AST. As such, the licensee is in compliance with the design requirements stated Inits UFSAR.

In producing an integrated assessment of the proposed change, the NRC staff also considered, on a qualitative basis, the overall Increase Inrisk. The principal DBA to be considered is the MHA that could pressurize the containment and be the driving force for leakage from the containment. The MHA has a very low frequency of occurrence. The MHA assumption that the entire core will be damaged is conservative. Assumptions about plate out of iodine and retention of iodine in containment fluids are also conservative. Core damage frequency and large early release fractions are unaffected by the PRVS. The resulting dose calculations show that the release would be within the licensing regulatory limits with substantial margins. These considerations provide some assurance that the uncertainties Inparameters and the progression of the accident will not induce a situation where the regulatory lirhits would be exceeded.

The NRC staff has concluded that the proposed change has no appreciable impact on plant safety. The proposed change will enhance efficiency of operation and reduce unnecessary regulatory burden by eliminating TS requirements for this system. Public confidence is maintained by carefully assessing the risk associated with this proposed change and ensuring that the impact on public health and safety is minimal and that any potential increase in exposure is conservatively within established regulatory guidelines.

The NRC staff finds that the removal of the PRVS from the TS is acceptable. Although there Is an increase in the unfiltered leakage from an accident and a reduction in defense-in-depth and the ability to mitigate increased leakage, the AST analysis shows acceptable dose results below the limits established in 10 CFR 50.67 and GDC 19. Aqualitative assessment of the increased risk Indicates that there are reasonable safety margins and that the Increased risk is not significant. The licensee has committed to the completion of modifications that support this change Including the change in control room filter efficiency testing, the change in containment Integrated leak rate test requirements, the use of dual air intakes on the control room, and the modification to limit ECCS leakage during post-LOCA recirculation. Each of these modifications is being implemented under the 10 CFR 50.59 process consistent with the

existing plant licensing basis. This finding of acceptability Is contingent upon these modifications being completed prior to the Implementation of the changes that have been found to be acceptable in this SE.

4.3 TS 3.7.17 Spent Fuel Pool Ventilation System

a. The licensee proposed to delete two notes in LCO 3.7.17:

'Note 1. LCO 3.0.3 Is not applicable Note 2. Not applicable during reracking operations with no fuel in the spent fuel pool.

In the February 18, 2004, letter, the licensee stated that during the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in which fuel Is considered to be recently Irradiated and the TS is applicable, there are no actions required by entry In 1CO 3.0.3, and thus, Note 1 serves no function.

The licensee also states that Note 2 Is unnecessary since during the short 72-hour period In which the TS Is applicable there would be no reracking operations and thus it doesn't serve a function. The NRC staff concurs that these notes may be deleted without Impact on public health and safety.

b. The licensee proposed changing the APPLICABILITY from during movement of fuel in the spent fuel poor to 'during movement of recently Irradiated fuel assemblies In the spent fuel pool."

The proposed change Is consistent with TSTF-51, which recognizes that the OPEIRABILITY of some systems can be relaxed after a sufficient period of time for fuel decay has passed. The period of time defined by 'recently" must be equal to or longer than the time used in the FHA design-basis analysis for decay of fuel.

The licensee used a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in its FHA DBA for fuel decay. The TS Bases are being revised to provide a definition of "recently irradiated" fuel. The NRC staff finds this change Is acceptable.

c. The licensee proposed deleting the statement 'During crane operations with loads over the spent fuel poor from the APPLICABILITY.

In the February 18, 2004, letter, the licensee stated that this requirement is being deleted since the revised FHA does not take credit for the SFPVS in the event of a Spent Fuel Pool Building Accident. The licensee also stated that for the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

• after shutdown In which the SFPVS is required to be operational, fuel can not be physically moved. The NRC staff concurs that crane operations with loads over the SFP for times outside of the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown pose no threat to public health and safety because the consequences of an FHA, after a decay time of at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, are well below regulatory limits.

d. The licensee proposed modifying the ACTIONS in CONDITION A and CONDITION B by changing the wording in the REQUIRED ACTION from 'Suspend movement of fuel in the spent fuel pool" to "Suspend movement of recently irradiated fuel assemblies in the spent fuel pool."

The proposed change Is consistent with TSTF-51, which recognizes that the OPERABILITY of some systems can be relaxed after a sufficient period of time for fuel decay has passed. The period of time defined by recently must be equal to or onger than the time used In the fuel-handling accident design-basis analysis for decay of fuel. The licensee used a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> In Its FHA DBA for fuel decay. The TS Bases are being revised to provide a definition of "recently irradiated" fuel. The NRC staff finds this change Is acceptable.

e. The licensee proposed modifying the ACTIONS in CONDITION A and CONDITION B by deleting the REQUIRED ACTION to "Suspend crane operations with loads over the spent fuel pool."

In the February 18, 2004, letter, the licensee stated that this requirement is being deleted since the revised FHA does not take credit for the SFPVS in the event of a Spent Fuel Pool Building Accident. The licensee also stated that for the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown in which the SFPVS Is required to be operational, fuel can not be physically moved. The NRC staff concurs that crane operations with loads over the SFP for times outside of the Initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown pose no threat to public health and safety because the consequences of an FH-A, after a decay time of at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, are well below regulatory limits.

f. The licensee proposed changing SURVEILLANCE REQUIREMENT SR 3.7.17.1 FREQUENCY from "31 days" to "Within 2,1 days prior to the movement of recently irradiated fuel assemblies."

The NRC staff concurs that the planned testing of the SFPVS train in the period of time of less than 31 days prior to the movement of irradiated fuel provides assurance that the system will be functioning and available, i.e., OPERABLE.

Although testing at other times may be useful as part of an overall maintenance or surveillance program, the NRC staff concludes that the testing during the immediate period before.moving -recently irradiated fuel is sufficient for the TS surveillance requirement.

4.4 TS 5.5.2 Containment Leakage Rate Testing Program

a. The licensee proposed changing the maximum allowable containment leakage rate, , at P., from 0.25% to 0.20% of containment air weight per day.

The NRC staff reviewed the licensee's request to change the leakage rate criteria from 025% per day to 0.20% per day. This is a conservative change in the sense that it allows a smaller leakage rate from the containment. It Is a non-conservative change in that it justifies a smaller assumed containment leakage rate in the MHA DBA with the overall effect of reducing the calculated dose. However, a review of test results from the licensee's most recent containment leakage testing shows that the licensee typically has containment leakages less than the 0.20% value that has been proposed. Thus, the change proposed by the licensee Is reasonable based on operating experience and the NRC staff finds that the proposed change to the more restrictive leakage rate test criteria is acceptable.

-N

b. The iensee proposed to delete the criterion Item b.that states Leakage >0.5 L, shall be to the penetration room.

The licensee does not take credit for containment leakage to the penetration room and its filter systern Inits MiHIA. The licensee assumes that all of the containment leakage, L., Is released directly to the environment. This Is conservative and the NRC staff finds that this change Is acceptable.

4.5 TS 5.5.12 Ventilation Filter Testing Program

a. TS 5.5.12 Items a and c.

The licensee proposed deleting the PRVS HEPA filter test Initem a and the PRVS carbon adsorber test InItem c.

As discussed previously, the PRVS system is not credited In any DBA. As such, the NRC staff has determined that the system may be removed from the TS.

Similarly, test requirements for the PRVS may also be removed. Requirements for testing this system will be controlled by other plant documents, if applicable. The NRC staff finds this change acceptable.

b. TS 5.5.12 item e.

The licensee proposed to delete reference to the PRVS and to change the laboratory test acceptance criteria for the CRVS booster fan filter from 90% to 97.5% and states that the criteria for the SFPVS carbon adsorber would remain at g90%.

Deletion of the reference to the PRVS is acceptable based on the deletion of the PRVS from the TS as discussed above. Increasing the acceptance criteria for the CRVS booster fan carbon adsorber is conservative and allows the licensee to take credit for a higher adsorption efficiency Inits DBA analysis. Leaving the SFPVS carbon adsorber efficiency at its current value of 290% does not adversely impact the operation of the system. The NRC staff finds these changes acceptable.

d. TS 5.5.12 item f.

The licensee proposed deleting this item since It pertains to the PRVS system which isbeing deleted from the TS.

The NRC staff finds that the deletion of this Item is acceptable since the PRVS system is being removed from the TS. Operational and testing requirements, if applicable, more appropriately belong in other plant documents.

4.6 TSTF-51 Related Technical Specification Changes Changes related to TSTF-51 establish a point in time following reactor shutdown when OPERABILITY of ESF systems that are typically used to mitigate the consequences of a FHA are no longer required to be OPERABLE in order to meet-the SRP guidance for offsite dose limits (less than 25 percent of 10 CFR Part 100 limits or the limits specified in 10 CFR 50.67).

S.'

Fuel that has been part of a critical reactor core within the preceding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Is referred to as "recently" irradicated fuel. Only recently irradiated fuel contains sufficient fission products to require OPERABILITY of an accident mitigation feature to meet the accident analysis assumptions. Therefore, the APPLICABILITY requirements for OPERABILITY FHA consequence mitigation features are revised. The requested changes would eliminate TS requirements for these features during core alterations involvng fuel movement beyond the recently Irradiated time period. The affected TS Umiting Conditions for Operation (LCO) are defined in the following TS sections:

TS Section Title Extent of change Number TS 3S.16 Reactor Building Deletes APPLICABILITY during CORE ALTERATIONS Applicability, (RB) Purge Adds recently In front of irradiated fuel Required Isolation - High Deletes Required Action A.2.1 Action, and Radiation Adds recently In front of irradiated fuel In Required SR 3.3.16.2 Action A.2.2 and renumbers It A.2 Deletes CORE ALTERATIONS and adds recently in front of irradiated fuel in the Surveillance Requirement TS 3.7.9 Control Room Adds APPLICABIUTY 'during movement of recently Ventilation System irradiated fuel assemblies."

(CRVS) Booster Changes Condition D to limit applicability for Modes Fans 1 2,3, or 4.

Adds Condftion'E for Applicability "during movement of recently irradiated fuel" along with Required Action "suspend movement and Completion Time

_Immediately" Table 3.7.16 Control Room Area Adds APPLICABILITY 'during movement of recently C6ooling Systems irradiated fuel assemblies."

{QRACS) Changes Condition D to limit applicability for Modes 1,2,3, or 4.

Renumbers Condition Eto Condition F and limits applicability for Modes 1,2,3, or 4.

Adds new Condition E for 'required action and completion time not met with Required Actions and Completion Times.

Adds new Condition G for 1wo CRACS trains inoperable during movement of recently irradiated fuel with Required action and Completion Time.

TS 3.B.2 AC Sources - Adds recently in front of irradiated fuel Applicability Shutdown and Actions A B and

TS Section TIe Extent of change Number TS 3.8A DC Sburces - Adds recently in front of irradiated fuel Applicability Shutdown and Required Action A.2.2 TS 3.8.7 Vital Inverters - Adds recently Infront of irradiated fuel Applicability Shutdown and Required Acti6n A.2.2 TS 3.8.9 Distribution Adds recently Infront of Irradiated fuel Applicability Systems -

and Shutdown Required Action A.2.2 TS 3:9.3 Containment Deletes APPLICABILITY during CORE ALTERATIONS Applicability, Penetrations Adds recently in front of Irradiated fuel Required Deletes Required Action A.1 Action, and Adds recently htfront of Irradiated fuel In Required SR 3.9.3.2 Action A.2 and renumbers it A.1 Deletes CORE ALTERATIONS and adds recently in front of irradiated fuel inthe Surveillance Requirement TS 3.9.6 Fuel Transfer Deletes APPLICABILITY 'during CORE Applicability Canal Water Level ALTERATIONS, except during latching and unlatching and of CONTROL ROD drive shafts" Required Deletes Required Action A.1 Action In order to implement the above changes to APPLICABILITY and ACTION statements, the LCO for OPERABILITY need only apply when handling fuel that has recently been in the critical reactor core (i.e., "recently irradiated" fuel). The submittal proposes a revised FHA dose analysis for Oconee which takes credit for a radioactive decay period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> based on an AST pursuant to 10 CFR 50.67 and the guidance of RG 1.183. Given this decay period, the licensee is now proposing changes to redefine the TS requirements by requiring ESF systems, which are previously relied upon to mitigate an FHA, applicable only during movement of fuel that has been "recently Irradiated." The term "recently Irradiated" is a cycle-specific number and represents the decay period for the reduction in radionuclide inventory available for release in the event of an FHA. For the upcoming refueling outage, the licensee has determined that the appropriate decay period will be 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In summary, once the reactor has been shutdown for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the licensee has demonstrated that the FHA reanalysis

(that does not rely on either building Integrity or the FHA mitigating systems) will not exceed dose limitations. The TS Bases are being revised to provide a definition of "recently Irradiated" fuel.

In addition, consistent with the Instructions in TSTF-51, Revision 2, regarding decreasing doses even further below that provided by natural decay, the licensee has committed to follow the guidelines of NUMARC 93-01, Revision 3, Section 11.3.6, Assessment Methods for Shutdown Conditions,* Subsection 5, Containment - Primary (PWR)/Secondary (BWR)."

The deletion of the CORE ALTERATIONS term Is justified since an FHA is the only event during CORE ALTERATIONS that Is postulated to result In fuel damage and radiological release, and such FHAs will be fully enveloped by the proposed APPLICABILITY.

In addition to the above changes to the APPLICABILITY statements, the licensee proposed numerous corresponding changes to the ACTION statements, such as elimination of references to CORE ALTERATIONS and the insertion of T recently Irradiated fuer when referring to the movement of irradiated fuel. The proposed changes do not Impact TS requirements for systems needed to prevent or mitigate CORE ALTERATION events other than the FHA. They also do not change the requirements for systems needed for decay heat removal or requirements to maintain the specified water levels over irradiated fuel. Since the proposed revisions to the TS do not result in changes to the design-basis, the NRC staff concludes that these revisions are acceptable.

4.7 IS 3.7.9 Control Room Ventilation System (CRVS) Booster Fans In addition to the TSTF-51 changes shown in the table above, the licensee proposes to add a Note to the Completion Time for Condition B and Condition C.

Condition B Completion Time additional note:

Note: An additional 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> is allowed when entering this condition for implementation of Control Room Intakelboosterfan modification.

Condition C Completion Time additional note:

Note: An additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is allowed when entering this condition for implementation of Control Room intake/booster fan modification The licensee states that the Notes will allow for a one-time additional completion time extension to Implement the Control Room Intake/Booster Fan modification. This extension is acceptable based on the current knowledge and experience in control room habitability. The completion times specified ir the proposed required actions recognize the low probability of an accident occurring during the time period when the boundary is degraded.

The NRC staff has reviewed the inclusion of these two notes and finds that they are acceptable for a one-time modification of the control room Intake/booster fans.

In summary, the NRC staff has reviewed the licensee's analyses and performed confirmatory assessments of the radiological consequence of the postulated LOCA and FHA. The doses

II calculated by the licensee are listed InTable 1. The doses calculated by the licensee and by the NRC staff are all within relevant dose criteria specified In 10 CFR 50.67 and the SRP.

Therefore, the NRC staff concludes that the radiological consequences analyzed and submitted by the licensee are acceptable predicated on a determination that both dual Intake channels for each Individual control room would perform continuously and simultaneously to Individually provide the air flow assumed In the licensee's analysis and that air from at least one of the two intakes could be assumed to be uncontaminated for the duration of an accident. The NRC staff further concludes that the proposed TS changes are acceptable and do not pose any significant risk to the health and safety of the public and control room operators.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments Involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsie and that there Is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously Issued proposed findings that the amendments Involve NSHC, and there has been r6o public comment on such findings (67 FR 2922, 68 FR 59215, and 68 FR 68660). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental Impact statement or environmental assessment need be prepared in connection with the Issuance of the amendments.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1)there is reasonable assurance that the health and safety of the public will not be endangered by opera~on in the proposed manner, (2) such activities will be conducted Incompliance with the Commission's regulations, and (3)the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Lee L. Brown E. Forrest Date: June 1, 2004

Table 1 Radiological Consequences Expressed as TEDEM' (rem)

Design Basis Accidents EABm LPZe Control Room LOCA 11.8 3.3 4.4 Dose criteria"') 25 25 5.0 Fuel handing accident for single fuel Assembly event 1.2 0.1 2.2 Dose criteria 6.3X5) 6.3 5) 5.0 4)

Fuel cask handing accident for multiple fuel Assembly event 1.8 02 3.4 Dose criteria 6 3(s) 6.30) 5-0(4) i"Submittal dated November 21, 2002 (Attachments 3 and 4)

') Exclusion area boundary

'3 Low population zone (4) 10 CFR 50.67

5) SRP 15.0.1

t..

Table 2 Parameters and Assumptions Used In Radiological Consequence Calculations Loss-of-Coolant Accident Parameter Value Reactor power 2568 MWt Containment volume 1.78E+6 ft3 Sprayed area 8.66E+5 ft3 Unsprayed area .9.17E+5 ft3 Containment leak rates 0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.2% per day 24 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.1% per day Containment mixing rates Sprayed to unsprayed 30,567 cfm Unsprayed to sprayed 30,567 cfm Aerosol removal rates by containment spray (per hour)

Time Rat es 0 to 96 seconds 0 96 seconds to 25 minutes 9.7 25 minutes to 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 6.7 3.5 to 112.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.6; 7 Elemental Iodine removal rates by spray (per hour)

Time Rati es 0 to 96 seconds 0 96 seconds to 25 minutes 20 25 minutes to 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0 1.8 to B hours 0.06 8 to 13.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 0.09 13.8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.13 24 to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.07 96 to 112.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> O.OC 72

r-.

Table 2 Parameters and AssumptlOns Used In Calculations Radiological Consequence Accident Loss of-Coolanth (Continued) 3 4.81 E+4 t Containment sump volume ECmS leak rates o

0 to 25 minutes 50 gph 25 minutes to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 10%

iodine partition factor ECCS leak rates to BWST Rates Time 0

0 to 25 minutes 5 gpm 25 minutes to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Control room 3 8.64E+4 f1 Volume 1150cfm Unfiltered air inleakage rates 40 cfm

. Oto 30 minutes 40 cfm 30 mninutes to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Filter efficiencies 99%

Aerosol iodine Elemental 95%

Organic iodine

Table 3 Parameters and Assumptions Used In Radiological Consequence Calculations Fuel Handling Accident Parameter Value Reactor power 2568 MWt Radial peaking factor Single fuel assembly event 1.65 Multiple fuel assembly cask drop event 1.20 Fission product decay period Single fuel assembly event 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Multiple fuel assembly cask drop event 55 days to 1 year Number of fuel assembly damaged Single fuel assembly event One Multiple fuel assembly cask drop event 518 to 1,024 Fuel pool water depth 21.34 ft Fuel gap fission product inventory Noble gases excluding Kr-85 5%

Kr-85 10%

1-131 8%

Alkali metals 12%

Fuel pool decontamination factors Iodine 183 Noble gases 1 Duration of accident 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

Table 4 Meteorological Data used for Loss-of-Coolant (LOCA) and Fuel Handling (FHA) Accidents Off site Relative Concentration (X/Q) Values for LOCA and FHA Time Mhr) ReceDtor X/Q (sec/r 3 )

0-2 EAB 2.2 E-4 O to8 LPZ 2.35 E-5 8 to 24 LPZ 4.70 E-6 24 to 96 LPZ 1.50 E-6 96 to 720 LPZ 3.30 E-7 Control Room Relative Concentration (X/Q) Values Time (hrl Accident X/Q(sec/r 3 )

0 to 2 LOCA (unit vent release) 4.79 E-4*

2 to 8 LOCA (unit vent release) 3.40 E-4*

8 to 24 LOCA (unit vent release) 1.40 E-4*

24 to 96 LOCA (unit vent release) 1.09 E-4*

96 to 720 LOCA (unit vent release) 8.86 E-5*

o to 2 LOCA (Borated water storage tank (BWST)) 2.13 E-4*

2 to 8 LOCA (BWST release) 1.61 E-4*

8 to 24 LOCA (BWST release) 6.66 E-5*

24 to 96 LOCA (BWST release) 5.19 E-5*

96 to 720 LOCA (BWST release) 4.06 E-5*

0 to 2 FHA 5.38 E-4**

2 to8 FHA 3.74 E-4**

8 to24 FHA 1.57 E-4**

24 to 96 FHA 1.24 E-4**

96 to 720 FHA 1.01 E-4**

• LOCA (unit vent release) and LOCA (BWST release) control room XJO estimates are based upon an assumption that wird flow Is In the direction of the Intake with the poorer atmospheric dispersion conditions and that this intake is drawinglIn 55 percent of the air.

FHAcontrol room X/C estimates are based upon an assumption that wind flow IsInthe direction of the intake with the pooer atmospheric dispersion conditions and that this intake isdrawing In60 percent of the air. Further, the intakes are assumed to be 10 feet higher than for the LOCA and BWST control room estimates. These assumptions were made as part of a 'sensitivity" study the licensee performed. Duke has stated that the final FHA doses calculated after the intakes are installed will reflect the as-tested control room air inflow and are expected to be lower than the dose values reported in Table 1.

Table 5 Measured Unfiltered Air Inleakage Rates 1998 Test Control Room Values Unit I and 2 80 +1- 55 cfm Unit 3 73+/- 25 cfm 2001 Test Control Room Values Unit 1 and 2 0 +/- 18 cfm Unit 3 o +/- 13 cfm