Text

Technical Specification Bases Changes
	ML081270228
Person / Time
Site:	Catawba
Issue date:	04/29/2008
From:	Morris J; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML081270228 (87)
	v • d • e

Duke JAMES R. MORRIS, VICE PRESIDENT tEEnergy, Duke Energy Carolinas,LLC Carolinas Catawba Nuclear Station CNO1 VP 4800 Concord Road York, SC 29745 803-831-4251 803-831-3221 fax April 29, 2008, U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Technical Specification Bases Changes Pursuant to 10CFR 50.4, please find attached changes to the Catawba Nuclear Station Technical Specification Bases. These Bases changes were made according to the provisions of 10CFR 50.59 and submitted on a frequency consistent with 10 CFR 50.71(e).

Any questions regarding this information should be directed to Marc Sawicki, Regulatory Compliance, at (803) 701-5191.

I certify that I am a duly authorized officer of Duke Energy Corporation and that the information contained herein accurately represents changes made to the Technical Specification Bases since the previous submittal.

James. R. Morris Attachment www. duke-energy. corn

U.S. Nuclear Regulatory Commission April 29, 2008, Page 2 Xc: V. M. McCree, Acting Regional Administrator Region II U.S. Nuclear Regulatory Commission Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 J. F. Stang, Jr., NRR Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 A.T. Sabisch Senior Resident Inspector Catawba Nuclear Station

U.S. Nuclear Regulatory Commission April 29, 2008, Page 3 bxc: w/o attachment NCMPA-1 NCEMC SREC PMPA w/attachment Electronic Licensing Library EC050 RGC File CN01RC Master File CN-801.01 CN04DM

DUKE ENERGY CORPORATION Phiukeq Catawba Nuclear Station 4800 Concord Road York, SC 29745 803 831 3000 April 28, 2008 Re: Catawba Nuclear Station Technical Specifications Bases Please replace the corresponding pages in your copy of the Catawba Technical Specifications Manual as follows:

REMOVE THESE PAGES INSERT THESE PAGES LIST OF EFFECTIVE PAGES Pages 1-33 Pages 1-33 TAB 3.1.1 B 3.1.1-3 thru B 3.1.1-6 B 3.1.1-3 thru B 3.1.1-6 TAB 3.3.1 B 3.3.1-1 thru B 3.3.1-2 B 3.3.1-1 thru B 3.3.1-2 B 3.3.1-51 B 3.1.1-51 TAB 3.3.6 B 3.3.6-1 thru B 3.3.6-2 B 3.3.6-1 thru B 3.3.6-2 B 3.3.6-5 B 3.3.6-5 TAB 3.4.13 B 3.4.13-1 thru B 3.4.13-2 B 3.4.13-1 thru B 3.4.13-2 B 3.4.13-5 thru B 3.4.13-6 B 3.4.13-5 thru B 3.4.13-6 Added Page B 3.4.13-7 TAB 3.4.16 B 3.4.16-1 thru B 3.4.16 -6 B 3.4.16-1 thru B 3.4.16 -6 www. duke-energy. corn

TAB 3.4.18 B 3.4.18-3 thru B 3.4.18-4 B 3.4.18-3 thru B 3.4.18-4 B 3.4.18-7 thru B 3.4.18-8 B 3.4.18-7 thru B 3.4.18-8 TAB 3.6.17 B 3.6.17-1 thru B 3.6.17-2 B 3.6.17-1 thru B 3.6.17-2 B 3.6.17-5 B 3.6.17-5 TAB 3.7.2 B 3.7.2-3 thru B 3.7.2-5 B 3.7.2-3 thru B 3.7.2-5 TAB 3.7.6 B 3.7.6-3 B 3.7.6-3 TAB 3.7.10 B 3.7.10-1 thru B 3.7.10-4 B 3.7.10-1 thru B 3.7.10-4 TAB 3.7.14 B 3.7.14-1 thru B 3.7.14-3 B 3.7.14-1 thru B 3.7.14-3 TAB 3.7.17 B 3.7.17-1 thru B 3.7.17-3 B 3.7.17-1 thru B 3.7.17-3 TAB 3.9.3 B 3.9.3-1 thru B 3.9.3-4 B 3.9.3-1 thru B 3.9.3-4 TAB 3.9.6 B 3.9.6-1 thru B 3.9.6-3 B 3.9.6-1 thru B 3.9.6-3 If you have any questions concerning the contents of this Technical Specification update, contact Betty Aldridge at (803)701-3758.

Randy Hart Manager, Regulatory Compliance

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Catawba Nuclear Station Technical Specifications List of Effective Pages Page Number Amendment Revision Date i 4/08/99 ii 219/214 3/01/05 iii 215/209 6/21/04 iv 173/165 9/30/98 1.1-1 173/165 9/30/98 1.1-2 173/165 9/30/98 1.1-3 197/190 4/22/02 1.1-4 179/171 8/13/99 1.1-5 197/190 4/22/02 1.1-6 *179/171 8/13/99 1.1.7 179/171 8/13/99 1.2-1 173/165 9/30/98 1.2-2 173/165 9/30/98 1.2-3 173/165 9/30/98 1.3-1 173/165 9/30/98 1.3-2 173/165 9/30/98 1.3-3 173/165 9/30/98 1.3-4 173/165 9/30/98 1.3-5 173/165 9/30/98 1.3-6 173/165 9/30/98 1.3-7 173/165 9/30/98 1.3-8 173/165 9/30/98 1.3-9 173/165 9/30/98 1.3-10 173/165 9/30/98 1.3-11 173/165 9/30/98 1.3-12 173/165 9/30/98 1.3-13 173/165 9/30/98 1.4-1 173/165 9/30/98 1.4-2 173/165 9/30/98 1.4-3 173/165 9/30/98 1.4-4 173/165 9/30/98 Catawba Units 1 and 2 Page 1 3/19/08

Page Number Amendment Revision Date 2.0-1 210/204 12/19/03 3.0-1 235/231 3/19/07 3.0-2 235/231 3/19/0.7 3.0-3 235/231 3/19/07

.3.0-4 235/231 3/19/07 3.0-5 235/231 3/19/07 3.0-6 235/231 3/19/07 3.1.1-1 173/1 65 9/30/98 3.1.2-1 173/1 65 9/30/98 3.1.2-2 173/1 65 9/30/98 3.1.3-1 173/1 65 9/30/98 3.1.3-2 173/1 65 9/30/98 3.1.3-3 173/1 65 9/30/98 3.1.4-1 173/1 65 9/30/98 3.1.4-2 173/1 65 9/30/98 3.1.4-3 173/1 65 9/30/98 3.1.4-4 173/1 65 9/30/98 173/1 65 9/30/98 3.1.5-1 173/1 65 9/30/98 3.1.5-2 173/1 65 9/30/98 3.1.6-1 173/1 65 9/30/98 3.1.6-2 173/1 65 9/30/98 3.1.6-3 173/1 65 9/30/98 3.1.7-1 173/1 65 9/30/98 3.1.7-2 173/1 65 9/30/98 3.1.8-1 173/1 65 9/30/98 3.218-2 173/1 65 9/30/98 3.2.1-1 173/1 65 9/30/98 3.2.1-2 173/1 65 9/30/98 3.2.1-3 180/1 72 9/22/99 3.2.1-4 180/1 72 9/22/99 Catawba Units 1 and 2 Pg 2 Page /90 3/19/08

Page Number Amendment Revision Date 3.2.2-1 173/165 9/30/98 3.2.2-2 173/165 9/30/98 3.2.2-3 173/165 9/30/98 3.2.2-4 180/172 9/22/99 3.2.3-1 173/165 9/30/98 3.2.4-1 173/165 9/30/98 3.2.4-2 173/165 9/30/98 3.2.4-3 173/165 9/30/98 3.2.4-4 173/165 9/30/98 3.3.1-1 173/165 9/30/98 3.3.1-2 173/165 9/30/98 3.3.1-3 207/201 7/29/03 3.3.1-4 207/201 7/29/03 3.3.1-5 207/201 7/29/03 3.3.1-6 173/165 9/30/98 3.3.1-7 173/165 9/30/98 3.3.1-8 173/165 9/30/98 3.3.1-9 173/165 9/30/98 3.3.1-10 173/165 9/30/98 3.3.1-11 173/165 9/30/98 3.3.1-12 173/165 9/30/98 3.3.1-13 173/165 9/30/98 3.3.1-14 179/171 8/13/99 3.3.1-15 179/171 8/13/99 3.3.1-16 179/171 8/13/99 3.3.1-17 179/171 8/13/99 3.3.1-18 210/204 12/19/03 3.3.1-19 210/204 12/19/03 3.3.1-20 173/165 9/30/98 3.3.2-1 173/165 9/30/98 3.3.2-2 173/165 9/30/98 3.3.2-3 173/165 9/30/98 Catawba Units 1 and 2 Page 3 3/19/08

Page Number Amendment Revision Date 3.3.2-4 173/1 65 9/30/98 3.3.2-5 173/1 65 9/30/98 3.3.2-6 173/1 65 9/30/98 3.3.2-7 181/1 73 11/02/99 3.3.2-8 181/1 73 11/02/99 3.3.2-9 224/219 5/24/05 3.3.2-10 2 08/2 02 9/10/03 3.3.2-11 196/1 89 3/20/02 3.3.2-12 179/1 71 8/13/99 3.3.2-13 208/2 02 9/10/03 3.3.2-14 20 8/2 02 9/10/03 3.3.2-15 214/208 5/12/04 3.3.3-1 219/214 3/1/05 3.3.3-2 219/214 3/1/05 3.3.3-3 219/214 3/1/05 3.3.3-4 219/214 3/11/05 3.3.4-1 213/207 4/29/04 3.3.4-2 173/1 65 9/30/98 3.3.4-3 173/1 65 9/30/98 3.3.5-1 173/1 65 9/30/98 3.3.5-2 179/1 71 8/13/99 3.3.6-1 196/1 89 3/20/02 3.3.6-2 224/219 5/24/05 3.3.6-3 196/1 89 3/20/02 3.3.7-1 (Deleted) 177/1 69 4/08/99 3.3.7-2 (Deleted) 177/1 69 4/08/99 3.3.7-3 (Deleted) 177/1 69 4/08/99 3.3.7-4 (Deleted) 177/1 69 4/08/99 3.3.8-1 (Deleted) 177/1 69 4/08/99 3.3.8-2 (Deleted) 177/1 69 4/08/99 3.3.8-3 (Deleted) 177/1 69 .4/08/99 3.3.9-1 207/201 7/29/03 Catawba Units 1 and 2 Pg 4 Page 3/19/08

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Page NumberAmnmnReionDtAmendment Revision Date 3.3.9-2 207/201 7/29/03 3.3.9-3 207/201 7/29/03 3.4.1-1 210/204 12/19/03 3.4.1-2 210/204 12/19/03 3.4.1-3 173/1 65 9/30/98 3.4.1-4 210/204 12/19/03 3.4.1-5 (deleted) 184/1 76 3/01/00 3.4.1-6 (deleted) 184/1 76 3/01/00 3.4.2-1 173/1 65 9/30/98 3.4.3-1 173/1 65 9/30/98 3.4.3-2 173/1 65 9/30/98 3.4.3-3 212/206 3/4/04 3.4.3-4 212/206 3/4/04 3.4.3-5 212/206 3/4/04 3.4.3-6 212/206 3/4/04 3.4.4-1 173/1 65 9/30/98 3.4.5-1 207/201 7/29/03 3.4.5-2 207/201 7/29/03 3.4.5-3 173/1 65 9/30/98 3.4.6-1 212/206 3/4./04 3.4.6-2 207/201 7/29/03 3.4.7-1 212/206 3/4/04 3.4.7-2 207/201 7/29/03 3.4.8-1 207/201 7/29/03 3.4.8-2 207/201 7/29/03 3.4.9-1 173/1 65 9/30/98 3.4.9 173/1 65 9/30/98 3.4.10-1 212/206 3/4/04 3.4.10-2 173/1 65 9/30/98 3.4-11-1 213/207 4/29/04 3.4.11-2 173/1 65 9/30/98 3.4.11-3 173/1 65 9/30/98 Catawba. Units 1 and 2 Pg 5 Page 3/19/08

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Page Number Amendment Revision Date 3.4.11-4 173/165 9/30/98 3.4.12-1 213/207 3/4/04 3.4.12-2 213/207 4/29/04 3.4.12-3 212/206 3/4/04 3.4.12-4 212/206 3/4/04 3.4.12-5 212/206 3/4/04 3.4.12-6 212/206 3/4/04 3.4.12-7 212/206 3/4/04 3.4.13-1 218/212 1/13/05 3.4.13-2 218/212 1/13/05 3.4.14-1 173/165 9/30/98 3.4.14-2 173/165 9/30/98 3.4.14-3 173/165 9/30/98 3.4.14-4 173/165 9/30/98 3.4.15-1 234/230 9/30/06 3.4.15-2 234/230 9/30/06 3.4.15-3 234/230 9/30/06 3.4.15-4 234/230 9/30/06 3:4.16-1 213/207 4/29/04 3.4.16-2 173/165 9/30/98 3.4.16-3 173/165 9/30/98 3.4.16-4 173/165 9/30/98 3.4.17-1 186/179 5/19/00 3.4.18-1 218/212 1/13/05 3.4.18-2 218/212 1/13/05 3.5.1-1 211/205 12/23/03 3.5.1-2 173/165 9/30/98 3.5.2-1 239/233 1/2/08 3.5.2-2 173/165 9/30/98 3.5.2-3 238/234 11/08/07 3.5.3-1 213/207 4/29/04 3.5.3-2 173/165 9/30/98 3.5.4-1 173/165 9/30/98 Catawba Units 1 and 2 Page 6 3/19/08

Page Number Amendment Revision Date 3.5.4-2 173/1 65 9/30/98 3.5.5-1 173/165 9/30/98 3.5.5-2 173/1 65 9/30/98 3.6.1-1 173/1 65 9/30/98 3.6.1-2 192/1 84 7/31/0 1 3.6.2-1 173/1 65 9/30/98 3.6.2-2 173/1 65 9/30/98 3.6.2-3 173/1 65 9/30/98 3.6.2-4 173/1 65 9/30/98 3.6.2-5 192/1 84 7/31/0 1 3..6.3-1 173/1 65 9/30/98 3.6.3-2 173/165 9/30/98 3.6.3-3 173/1 65 9/30/98 3.6.3-4 173/1 65 9/30/98 3.6.3-5 173/1 65 9/30/98

-3.6.3-6 225/220 6/10/05 3.6.3-7 192/1 84 7/31/01 3.6.4-1 173/1 65 9/30/98 3.6.5-1 .173/1 65 9/30/98 3.6.5-2 173/1 65 9/30/98 3.6.6-1 22 8/22 3 11/17/05 3.6.6-2 173/1 65 9/30/98 3.6.8-1 213/207 4/29/04 3.6.8-2 173/1 65 9/30/98 3.6.9-1 173/1 78 5/05/00 3.6.9-2 173/1 78 5/05/00 3.6.10-1 173/1 65 9/30/98 3.6.10-2 227/222 9/30/05 3.6.11-1 173/1 65 9/30/98 3.6.11-2 174/1 66 1/14/99 3.6.12-1 173/1 65 9/30/98 3.6.12-2 209/203 9/29/03 3.6.12-3 2 09/2 03 9/29/03 Catawba Units 1 and 2 Pg 7 Page /90 3/19/08

Page Number Amendment PageNumer AendentRevision Date 3.6.13-1 173/1 65 9/30/98, 3.6.13-2 173/1 65 9/30/98 3.6.13-3 173/1 65 9/30/98 3.6.14-1 173/1 65 9/30/98 3.6.14-2 173/1 65 9/30/98 3.6.14-3 173/1 65 9/30/98 3.6.15-1 173/1 65 9/30/98 3.6.15-2 173/1 65 9/30/98 3.6.16-1 178/1 70 4/09/99 3.6.16-2 227/222 9/30/05 3.6.17-1 22 8/22 3 11/17/05 3.7.1-1 173/1 65 9/30/98 3.7.1-2 173/1 65 9/30/98 3.7.1-3 173/1 65 9/30/98 3.7.2-1 173/1 65 9/30/98 3.7.2-2 173/1 65 9/30/98 3.7.3-1 173/1 65 9/30/98 3.7.3-2 173/1 65 9/30/98 3.7.4-1 213/207 4/29/04 3.7.4-2 173/1 65 9/30/98 3.7.5-1 22 8/22 3 11/17/05 3.7.5-2 173/1 65 9/30/98 3.7.5-3 173/1 65 9/30/98 3.7.5-4 173/1 65 9/30/98 3.7.6-1 173/1 65 9/30/98 3.7.6-2 173/1 65 9/30/98 3.7.7-1 228/223 11/17/05 3.7.7-2 173/1 65 9/30/98 3.7..8-1 22 8/22 3 11/17/05 3.7.8-2 173/1 65 9/30/98 3.7.9-1 232/228 9/25/06 3.7.10-1 228/223 11/17/05 Catawba Units 1 and 2 Pg Page 8 /90 3/19/08

Page Number Amendment Revision Date 3.7.10-2 198/19 1 4/23/02 3.7.10-3 187/1 80 9/05/00 3.7.11-1 198/191 4/23/02 3.7.11-2 198/19 1 4/23/02 3.7.12-1 239/223 1/2/08 3.7.12-2 173/1 65 9/30/98 3.7.13-1 198/19 1 4/23/02 3.7.13-2 176/1 68 3/26/99 3.7.14-1 173/1 65 9/30/98 3.7.15-1 173/1 65 9/30/98 3.7.16-1 233/229 9/27/06 3.7.16-2 233/229 9/27/06 3.7.16-3 233/229 9/27/06 3.7.17-1 173/1 65 9/30/98 3.8.1-1 22 8/22 3 11/17/05 3.8.1-2 173/1 65 9/30/98 3.8.1-3 228/223 11/17/05 3.8.1-4 173/1 65 9/30/98 3.8.1-5 173/1 65 9/30/98 3.8.1-6 173/1 65 9/30/98 3.8.1-7; 173/1 65 9/30/98 3.8.1-8 173/1 65 9/30/98 3.8.1-9 173/1 65 9/30/98 3.8.1-10 173/1 65 9/30/98 3.8.1-11 236/232 6/25/07 3.8.1-12 173/1 65 9/30/98 3.8.1-13 173/1 65 9/30/98 3.8.1-14 173/1 65 9/30/98 3.8.1-15 .173/1 65 9/30/98 3.8.2-1 173/1 65 9/30/98 3.8.2-2 207/201 7/29/03 Catawba Units 1 and 2 Pg 9 Page /90 3/19/08

Page. Number Amendment Revision Date 3.8.2-3 173/1 65 9/30/98 3.8.3-1 175/1 67 1/15/99 3.8.3-2 173/1 65 9/30/98 3.8.3-3 206/200 7/10/03 3.8.4-1 173/1 65 9/30/98 3.8.4-2 223/2,18 4/27/05 3.8.4-3 223/218 4/27/05 3.8.4-4 183/1 75 1/07/00 3.8.5-1 173/1 65 9/30/98 3.8.5-2 207/201 7/29/03 3.8.6-1 223/218 4/27/05 3.8.6-2 223/218 4/27/05 3.8.6&3 223/218 4/27/05 3.8.6-4 223/218 4/27/05 3.8.6-5 223/218 4/27/05 3.8.7-1 173/1 65 9/30/98 3.8.7-2 173/1 65 9/30/98 3.8.8-1 173/165 9/30/98 3.8.8-2 207/201 7/29/03 3.8.9-1 173/1 65 9/30/98 3.8.9-2 173/1 65 9/30/98 3.8.9-3 173/1 65 9/30/98 3.8.10-1 207/201 7/29/03 3..8.10-2 173/1 65 .9/30/98 3.9.1-1 226/221 9/1/05 3.9.2-1 215/209 6/21/04 3.9.2-2 215/209 -6/21/04 3.9.3-1 227/222 9/30/05 3.9.3-2 173/1 65 9/30/98 3.9.4-1 207/201 7/29/03 3.9.4-2 173/1 65 9/30/98 3.9.5-1 207/201 7/29/03 Catawba Units 1 and 2Pae13/90 Page 10 3/19/08

Page Number Amendment Revision Date 3.9.5-2 173/1 65 9/30/98 3.9.6-1 173/1 65 9/30/98 3.9.7-1 215/209 6/21/04 4.0-1 220/215 3/03/05 4.0-2 233/229 9/27/06 5.1-1 173/1 65 9/30/98 5.2-1 173/1 65 9/30/98 5.2-2 173/1 65 9/30/98 5.2-3 173/1 65 9/30/98 5.3-1 181/1 73 11/02/99 5.4-1 173/1 65 9/30/98 5.5-1 205/198 3/12/03 5.5-2 205/1 98 3/12/03 5.5-3 173/1 65 9/30/98 5.5-4 173/1 65 9/30/98 5.5-5 216/210 8/5/04 5.5-6 218/212 1/13/05 5.5-7 218/212 1/13/05 5.5-7a ---/233 10/31/07 5.5-8 218/212 1/13/05 5.5-9 218/212 1/13/05 5.5-10 22 7/22 2 9/30/05 5.5-1 1 227/222 9/30/05 5.5-12 218/2 12- 1/13/05 5.5-13 218/212 1/13/05 5.5-14 218/212 1/13/05 5.6-1 222/217 3/31/05 5.6-2 222/217 3/31/05 5.6-3 222/217 3/31/05 Catawba Units 1 and 2Pae13/90 Page 11 3/19/08

Page Number Amendment Revision Date 5.6-4 222/217- 3/31/05 5.6-5 222/217 3/31/05 5.7-1 173/1 65 9/30/98 5.7-2 173/1 65 9/30/98 BASES Revision 1 4/08/99 Revision 2 3/01/05 Revision 1 6/21/04 B 2. 1.1 -1 Revision 0 9/30/98 B 2.1.1-2 Revision 1 12/19/03 B 2.1.1-3 Revision 1 12/19/03 B 2.1.1-4 Revision 0 9/30/98 B 2.1.2-1 Revision 0 9/30/98 B 2.1.2-2 Revision 0 9/30/98 B 2.1.2-3 Revision 0 9/30/98 B 3.0-1 Revision'1 3/19/07 B 3.0-2 Revision 1 3/19/07 B 3.0-3 Revision 2 3/19/07 B 3.0-4 Revision 3 3/19/07 B 3.0-5 Revision 3 3/19/07 B 3.0-6 Revision 2 3/19/07 B 3.0-7 Revision 2 3/19/07 B 3.0-8 Revision 3 3/19/07 B 3.0-9 Revision 2 3/19/07 B 3.0-10 Revision 3 3/19/07 B 3.0-1 1 Revision 3 3/19/07 B 3*.0-12 Revision 3 3/19/07 B 3.0-13 Revision 3 3/19/07 B 3.0-14 Revision 3 3/19/07 B 3.0-15 Revision 1. 3/19/07 B 3.0-16 Revision 1 3/19/07 B 3.0-17 Revision 0 3/19/07 Catawba Units 1 and 2Pae13/90 Page 12 3/19/07

Page NumberAmn Amendment en Revision Date B 3.0-18 Revision 3/19/07 B 3.0-19 Revision 3/19/07 B 3. 1.1 -1 Revision 7/13/05 B 3.1.1-2 Revision 7/13/05 B 3.1.1-3 Revision 7/13/05 B 3.1.1-4 Revision 3/13/08 B 3.1.1-5 Revision 7/13/05 B 3.1.1-6 Revision 3/13/08 B 3.1.2-1 Revision 9/30/98 B 3.1.2-2 Revision 4/26/00 B 3.1.2-3 Revision 9/30/98 B 3.1.2-4 Revision 9/30/98 B 3.1.2-5 Revision 9/30/98 B 3.1.3-1 Revision 4/26/00 B 3.1.3-2 Revision 4/26/00 B 3.1 .3-3 Revision 4/26/00 B 3.1 .3-4 Revision 4/26/00 B 3.1.3-5 Revision 4/26/00 B 3.1 .3-6 Revision 4/26/00 B 3.1.4-1 Revision 9/30/98 B 3.1 .4-2 Revision 9/30/98 B 3.1.4-3 Revision .9/30/98 B 3.1.4-4 Revision 9/30/98 B 3.1 .4-5 Revision 9/30/98 0 3.1.4-6 Revision 9/30/98 B 3.1 .4-7 Revision 9/30/98 B 3.1.4-8 Revision 9/30/98 B 3.1.4-9 Revision 9/30/98 B 3.1.5-1 Revision B 3.1.5-2 Revision 2/18/02 B 3.1.5-3 Revision 9/30/98 B 3.1.5-4 Revision 9/30/98 Catawba Units 1 and 2Pae13/90 Page 13 3/19/08

Page Number Amendment Revision Date B 3.1.6-1 Revision 0 9/30/98 B 3.1.6-2 Revision 0 9/30/98 B 3.1.6-3 Revision 0 9/30/98 B 3.1 .6-4 Revision 0 9/30/98 B 3.1.6-5 Revision 0 9/30/98 B 3.1.6-6 Revision 0 9/30/98 B 3.1.7-1 Revision 0 9/30/98 B 3.1.7-2 Revision 2 1/08/04 B 3.1 .7-3 Revision 2 1/08/04 B 3.1 .7-4 Revision 2 1/08/04 B 3.1.7-5 Revision 2 1/08/04 B 3.1 .7-6 Revision 2 1/08/04 B 3.1.8-1 Revision 0 9/30/98 B 3.1.8-2 Revision 0 9/30/98 B 3.1.8-3 Revision 0 9/30/98 B 3.1.8-4 Revision 1 10/06/05

.B 3.1 .8-5 Revision 0 9/30/98 B 3.1.8-6 Revision 0 9/30/98, B 3.2.1 -1 Revision 0 9/30/98 B 3.2.1-2 Revision 1 10/02/00 B 3.2.1-3 Revision 1 10/06/05 B 3.2.1-4 Revision 1 10/06/05 B 3.2.1-5 Revision 1 10/02/00 B 3.2.1-6 Revision 0 9/30/98 B 3.2.1-7 Revision 0 9/30/98 B 3.2.1-'8 Revision 0 9/30/98 B 3.2.1-9 Revision 0 9/30/98 B 3.2.1 -10 Revision 0 9/30/98 B 3.2.1 -11 Revision 3 10/01/02 B 3.2-.2-1 Revision 1 3/01/00 B 3.2.2-2 Revision 2 10/02/00 B 3.2.2-3 Revision 1 3/01/00 B 3.2.2-4 Revision 1 3/01/00 Catawba Units 1 and 2Pae13/90 Page 14 3/19/08

Page Number Amendment, Revision Date B 3.2.2-5 Revision 1 3/01/00 B 3.2.2-6 Revision 1 3/01/00 B 3.2.2-7 Revision 1 3/01/00 B 3.2.2-8 Revision 1 3/01/00 B 3.2.2-9 Revision 2 3/01/00 B 3.2.2-10 Revision 1 3/01/00 B 3.2.3-1 Revision 0 9/30/98 B 3.2.3-2 Revision 0 9/30/98 B 3.2.3-3 Revision 0 9/30/98 B 3.2.3-4 Revision 1 10/01/02 B 3.2.4-1 Revision 1 10/02/00 B 3.2.4-2 Revision 1 2/ 26/99 B 3.2.4-3 Revision 0 9/30/98 B 3.2.4-4 Revision 0 9/30/98 B 3.2.4-5 Revision 1 2/26/99 B 3.2.4-6 Revision 1 11/5/03 B 3.2.4-7 Revision 1 11/5/03 B 3.3.1 -1 Revision 2 3/13/08 B 3.3.1-2 Revision 1 8/13/99 B 3.3.1-3 Revision 0 9/30/98 B 3.3. 1-4 Revision 1 8/13/99 B 3.3.1-5 Revision 0 9/30/98 B 3.3.1-6 Revision 0 9/30/98 B 3.3.1-7 Revision 0 9/30/98 B 3.3.1-8 Revision 0 9/30/98 B 3.3.1-9 Revision 0 9/30/98 B 3.3.1 -10 Revision 1 2/26/02 B 3.3.1 -11 Revision 0 9/30/98 B 3.3.1 -12 Revision 0 9/30/98 B 3.3.1-13 Revision 0 9/30/98 B 3.3.1-14 Revision 0 9/30/98 B 3.3.1-15 Revision 0 9/30/98 Catawba Units 1 and.2Pae13/90 Page 15 3/19/08,

Page Number Amendment Revision Date B 3.3.1-16 Revision 0 9/30/98 B 3.3.1-17 Revision 0 9/30/98 B 3.3.1-18 Revision 2 12/19/03 B 3.3.1-19 Revision 0 9/30/98 B 3.3.1-20 Revision 0 9/30/98 B 3.3.1-21 Revision 0 9/30/98 B 3.3.1-22 Revision 0 9/30/98 B 3.3.1-23 Revision 0 9/30/98 B 3.3.1-24 Revision 0 9/30/98 B 3.3.1-25 Revision 0 9/30/98 B 3.3.1-26 Revision 0 9/30/98 B 3.3.1-27 Revision 0 9/30/98 B 3.3.1-28 Revision 0 9/30/98 B 3.3.1-29 Revision 0 9/30/98 B 3.3.1-30 Revision 1 8/13/99 B 3.3.1-31 Revision 1 8/13/99 B 3.3.1-32 Revision 0 9/30/98 B 3.3.1-33 Revision 0 9/30/98 B 3.3.1-34 Revision 0 9/30/98 B 3.3.1-35 Revision 1 7/29/03 B 3.3.1-36 Revision 1 7/29/03 B 3.3.1-37 Revision 0' 9/30/98 B 3.3.1-38 Revision 0 9/30/98 B 3.3.1-39 Revision 0 9/30/98 B 3.3.1-40 Revision 0 9/30/98 B 3.3.1-41 Revision 0 9/30/98 B 3.3.1-42 Revision 0 9/30/98 B 3.3.1-43 Revision 0 9/30/98 B 3.3.1-44 Revision 0 9/30/98 B 3.3.1-45 Revision 2 6/13/05 B 3.3.1-46 Revision 2 6/13/05 B 3.3.1-47 Revision 1 6/13/05 Catawba Units 1 and 2 Page 16 3/19/08

Page Number Amendment Revision Date B 3.3.1-48 Revision 0 9/30/98 B 3.3.1-49 Revision 1 11/24/04 B 3.3.1-750 Revision 1 4/22/02 B 3.3.1-51 Revision 2 3/13/08 B 3.3.2-1 Revision 0 9/30/98 B 3.3.2-2 Revision 1 8/13/99 B 3.3.2-3 Revision 1 8/13/99 B 3.3.2-4 Revision 0 9/30/98 B 3.3.2-5 Revision 0 9/30/98 B 3.3.2-6 Revision 0 9/30/98 B 3.3.2-7 Revision 0 9/30/98 B 3.3.2-8 Revision 0 9/30/98 B 3.3.2-9 Revision 0 9/30/98 B 3.3.2-10 Revision 0 9/30/98 B 3.3.2-11 Revision 0 9/30/98 B 3.3.2-12 Revision 0 9/30/98 B 3.3.2-13 Revision 0 9/30/98 B 3.3.2-14 Revision 1 2/26/99 B 3.3.2-15 Revision 0 9/30/98 B 3.3.2-16 Revision 0 9/30/98 B 3.3.2-17 Revision 0 9/30/98 B 3.3.2-18 Revision 1 11/5/03 B'3.3.2-19 Revision 2 11/5/03 B 3.3.2-20 Revision 2 11/5/03 B 3.3.2-2 1 Revision 2 11/5/03 B 3.3.2-22 Revision 2 11/5/03 B 3.3.2-23 Revision 2 11/5/03 B 3.3.2-24 Revision 1 9/10/03 B 3.3.2-25 Revision 1 9/10/03 B 3.3.2-26 Revision 1 9/10/03 B 3.3.2-27 Revision 1 9/10/03

-B 3.3.2-28 Revision 1 9/10/03 Catawba Units 1 and 2Pae13/90 Page 17 3/19/08

Page Number Amendment Revision Date B 3.3.2-29 Revision 1 9/10/03 B 3.3.2-30 Revision 2 5/12/04 B 3.3.2-3 1 Revision 2 10/10/06 B 3.3.2-32 Revision 2 10/10/06 B 3.3.2-33 Revision 2 10/10/06 B 3.3.2-34 Revision 1 10/10/06 B 3.3.2-35 Revision 1 10/10/06 B 3.3.2-36 Revision 2 10/10/06 B 3.3.2-37 Revision 2 10/10/06 B 3.3.2-38 Revision 2 10/10/06 B 3.3.2-39 Revision 2 10/10/06 B 3.3.2-40 Revision 3 10/10/06 B 3.3.2-41 Revision 4 10/10/06 B 3.3.2-42 Revision 4 10/10/06 B 3.3.2-43 Revision 1 9/10/03 B 3.3.2-44 Revision 1 9/10/03 B 3.3.2-45 Revision 2 5/24/05 B 3.3.2-46 Revision 2 5/24/05 B 3.3.2-47 Revision 3 5/24/05 B 3.3.2-48 Revision 2 5/24/05 B 3.3.2-49 Revision 1 5/24/05 B 3.3.3-1 Revision 0 9/30/98 B 3.3.3-2 Revision 0 9/30/98 B 3.3.3-3 Revision 0 9/30/98 B 3.3.3-4 Revision 0 9/30/98 B 3.3.3-5 Revision 0 9/30/98 B 3.3.3-6 Revision 0 9/30/98 B 3.3.3-7 Revision 0 9/30/98 B 3.3.3-8 Revision 1 3/01/05 B 3.3.3-9 Revision 0 9/30/98 B 3.3.3-10 Revision 1 5/19/00 B 3.3.3-1 1 Revision 0 9/30/98 Catawba Units 1 and 2Pae13/90 Page 18 3/19/08

Page Number Amendment Revision Date B 3.3.3-12 Revision 2 3/01/05 B 3.3.3-13 Revision 0 9/30/98 B 3.3.3-14 Revision 2 3/01/05 B 3.3.3-15 Revision 1 3/01/05 B 3.3.3-16 Revision 0 9/30/98 B 3.3.4-1 Revision 0 9/30/98 B 3.3.4-2 Revision 0 9/30/98 B 3.3.4-3 Revision 1 4/29/04 B 3.3.4-4 Revision 0 9/30/98 B 3.3.4-5 Revision 0. 9/30/98 B 3.3.5-1 Revision 0 9/30/98 B 3.3.5-2 Revision 1 8/13/99 B 3.3.5-3 Revision 1 8/13/99 B 3.3.5-4 Revision 1 8/13/99 B 3.3.5-5 Revision 0 9/30/98 B 3.3.5-6 Revision 0 9/30/98 B 3.3.6-1 Revision 2 3/13/08 B 3.3.6-2 Revision 1 3/20/02 B 3.3.6-3 Revision 1 3/20/02 B 3.3.6-4 Revision 1 3/20/02 B 3.3.6-5 Revision 3 3/13/08 B 3.3.7-1 Deleted 4/08/99 B 3.3.7-2 Deleted 4/08/99 B 3.3.7-3 Deleted 4/08/99 B 3.3.7-4 Deleted 4/08/99 B 3.3.7-5 Deleted 4/08/99 B 3.3.8-1 Deleted 4/08/99 B 3.3.8-2 Deleted 4/08/99 B 3.3.8-3 Deleted 4/08/99 B 3.3.8-4 Deleted 4/08/99 B 3.3.9-1 Revision 0 9/30/98 B 3.3.9-2 Revision 1 7/29/03 Catawba Units 1 and 2Pae13/90 Page 19 3/19/08

Page Number Amendment Revision Date B 3.3.9-3 Revision 1 7/29/03 B 3.3.9-4 Revision 0 9/30/98 B 3.3.9-5 Revision 0 9/30/98 B 3.4.1 -1 Revision 2 12/19/03 B 3.4.1-2 Revision 2 12/19/03 B 3.4.1-3 Revision 2 12/19/03 B 3.4.1-4 Revision 2 12/19/03 B 3.4.1-5 Revision 0 9/30/98 B 3.4.2-1 Revision 0 9/30/98 B 3.4.2-2 Revision 0 9/30/98 B 3.4.2-3 Revision 0 9/30/98 B 3.4.3-1 Revision 1 3/4/04 B 3.4.3-2 Revision 1 3/4/04 B 3.4.3-3 Revision 1 3/4/04 B 3.4.3-4 Revision 1 3/4/04 B 3.4.3-5 Revision 1 3/4/04 B 3.4.3-6 Revision 1 3/4/04 B 3.4.4-1 Revision 0 9/30/98 B 3.4.4-2 Revision 1 1/13/05 B 3.4.4-3 Revision 0 9/30/98 B 3.4.5-1 Revision 0 9/30/98 B 3.4.5-2 Revision 0 9/30/98 B 3.4.5-3 Revision 2 1/13/05 B 3.4.5-4 Revision 1 4/26/00 B 3.4.5-5 Revision 2 7/29/03 B 3.4.5-6 Revision 2 7/29/03 B 3.4.6-1 Revision 0 9/30/98 B 3.4.6-2 Revision 3 1/13/05 B 3.4.6-3 Revision 0 9/30/98 B 3.4.6-4 Revision 1 7/29/03 B 3.4.6-5 Revision 1 7/29/03 B 3.4.7-1 Revision 0 9/30/98 Catawba Units 1 and 2Pae23/90 Page 20 3/19/08

Page Number Amendment Revision Date B 3.4.7-2 Revision 1 7/29/03 B 3.4.7-3 Revision 4 1/13/05 B 3.4.7-4 Revision 2 7/29/03 B 3.4.7-5 Revision 2 7/29/03 B 3.4.8-1 Revision 1 7/29/03 B 3.4.8-2 Revision 2 7/29/03 B 3.4.8-3 Revision 2 7/29/03 B 3.4.9-1 Revision 0 9/30/98 B 3.4.9-2 Revision 0 9/30/98 B 3.4.9-3 Revision 0 9/30/98 B 3.4.9-4 Revision 1 4/27/99 B 3.4.10-1 Revision 1 3/4/04 B 3.4.10-2 Revision 0 9/30/98 B 3.4.10-3 Revision 1 3/4/04 B 3.4.10-4 Revision 1 3/4/04 B 3.4.11 -1 Revision 0 9/30/98 B 3.4.11-2 Revision 1 11/5/03 B 3.4.11-3 Revision 3 4/29/04 B 3.4.11-4 Revision 1 11/5/03 B 3.4.11-5 Revision 0 9/30/98 B 3.4.11-6 Revision 0 9/30/98 B 3.4.11-7 Revision 0 9/30/98 B 3.4.12-1 Revision 1 3/4/04 B 3.4.12-2 Revision 1 3/4/04 B 3.4.12-3 Revision 1 3/4/04 B 3.4.12-4 Revision 1 3/4/04 B 3.4.12-5 Revision 1 3/4/04 B 3.4.12-6 Revision 1 3/4/04 B 3.4.12-7 Revision 1 3/4/04 B 3.4.12-8 Revision 2 4/29/04 B 3.4.12-9 Revision 2 4/29/04 B 3.4.12-10 Revision 2 4/29/04 Catawba Units 1 and 2Pae23/90 Page 21 3/19/08

Page Number Amendment Revision Date B 3.4.12-11 Revision 2 4/29/04 B 3.4.12-12 Revision 1 4/29/04 B 3.4.12-13 Revision 1 4/29/04 B 3.4.13-1 Revision 0 9/30/98 B 3.4.13-2 Revision 2 3/13/08 B 3.4.13-3 Revision 1 1/13/05 B 3.4.13-4 Revision 1 1/13/05 B 3.4.13-5 Revision 3 1/13/05 B 3.4.13-6 Revision 4 3/13/08 B-3.4.13-7 Revision 0 3/13/08 B 3.4.14-1 Revision 0 9/30/98 B 3.4.14-2 Revision 1 2/26/99 B 3.4.14-3 Revision 0 9/30/98 B 3.4.14-4 Revision 0 9/30/98 B 3.4.14-5 Revision 0 9/30/98 B 3.4.14-6 Revision 1 2/26/99 B 3.4.15-1 Revision 2 7/25/07 B 3.4.15-2 Revision 2 7/25/07 B 3.4.15-3 Revision 3 7/25/07 B 3.4.15-4 Revision 2 7/25/07 B 3.4.15-5 Revision 2 7/25/07 B 3.4.15-6 *Revision 4 7/25/07 B 3.4.15-7 Revision 1 7/25/07 B 3.4.15-8 Revision 1 7/25/07 B 3.4.15-9 Revision 1 7/25/07 B 3.4.15-10 Revision 0 7/25/07 B 3.4.16-1 Revision 1 3/13/08 B 3.4.16-2 Revision 1 3/13/08 B 3.4.16-3 Revision 2 3/13/08 B 3.4.16-4 Revision 2 3/13/08 B 3.4.16-5 Revision 2 3/13/08 B 3.4.16-6 Revision 1 3/13/08 B 3.4.17-1 Revision 1 1/13/05 B 3.4.17-2 Revision 0 9/30/98 Catawba Units 1 and 2g Page 22 3/19/08

Page Number IAmendment Revision Date B 3.4.17-3 Revision 1 5/19/00 B 3.4.18-1 Revision 0 1/13/05 B 3.4.18-2 Revision 0 1/13/05 B 3.4.18-3 Revision 1 3/13/08 B 3.4.18-4 Revision 0 1/13/05 B 3.4.18-5 Revision 0 1/13/05 B 3.4.18-6 Revision 0 1/13/05 B 3.4.18-7 Revision 0 1/13/05 B 3.4.18-8 Revision 1 3/13/08 B 3.5.1 -1 Revision 0 9/30/98 B 3.5.1-2 Revision 0 9/30/98 B 3.5.1-3 Revision 2 10/06/05 B 3.5.1-4 Revision 3 10/06/05 B 3.5.1-5 Revision 3 10/06/05 B 3.5.1-6 Revision 2 10/06/05 B 3.5.1-7 Revision 2 10/06/05 B 3.5.1-8 Revision 2 10/06/05 B 3.5.2-1 Revision 0 9/30/98 B 3.5.2-2 Revision 0 9/30/98 B 3.5.2-3 Revision 1 10/02/00 B 3.5.2-4 Revision 0 9/30/98 B 3.5.2-5 Revision 0 9/30/98 B 3.5.2-6 Revision 0 9/30/98 B 3.5.2-7 Revision 0 9/30/98 B 3.5.2-8 Revision 1 5/17/04 B 3.5.2-9 Revision 2 11/08/07 B 3.5.2-10 Revision 1 11/08/07 B 3.5.3-1 Revision 0 9/30/98 B 3.5.3-2 Revision 1 4/29/04 B 3.5.3-3 Revision 1 4/29/04 B 3.5.4-1 Revision 0 9/30/98 B 3.5.4-2 Revision 0 9/30/98 B 3.5.4-3 Revision 2 10/06/05 Catawba Units 1 and 2Pae23/90 . Page 23 3/19/08

Page Number Amendment Revision Date B 3.5.4-4 Revision 2 10/06/05 B 3.5.4-5 Revision 2 10/06/05 B 3.5.4-6 Revision 0 10/06/05 B 3.5.5-1 Revision 0 9/30/98 B 3.5.5-2 Revision 0 9/30/98 B 3.5.5-3 Revision 0 9/30/98 B 3.5.5-4 Revision 0 9/30/98 B 3.6.1 -1 Revision 1 7/31/01 B 3.6.1-2 Revision 1 7/31/01 B 3.6.1-3 Revision 1 7/31/01 B 3.6.1-4 Revision 1 7/31/01 B 3.6.1-5 Revision 1 7/31/0 1 B 3.6.2-1 Revision 0 9/30/98 B 3.6.2-2 Revision 1 7/31/01 B 3.6.2-3 Revision 1 7/31/01 B 3.6.2-4 Revision 1 7/31/01 B 3.6.2-5 Revision 1 7/31/01 B 3.6.2-6 Revision 1 7/31/01 B 3.6.2-7 Revision 1 7/31/01 B 3.6.2-8 Revision 1 7/31/01 B 3.6.3-1 Revision 1 3/20/02 B 31.6.3-2 Revision 1 3/20/02 B 3.6.3-3 Revision 1 3/20/02-B 3.6.3-4 Revision 1 3/20/02 B 3.6.3-5 Revision 0 9/30/98 B 3.6.3-6 Revision 0 9/30/98 B 3.6.3-7 Revision 0 9/30/98 B 3.6.3-8 Revision 0 9/30/98 B 3.6.3-9 Revision 0 9/30/98 B 3.6.3-10 Revision 0 9/30/98 B 3.6.3-1 1 Revision 0 .9/30/98 B 3.6.3-12 Revision 0 9/30/98 Catawba Units 1 and 2Pae23/90 Page 24 3/19/08

Page Number Amendment Revision Date B 3.6.3-13 Revision 3 12/05/05 B 3.6.3-14 Revision 1 7/3 1/0 1 B 3.6.3-15 Revision 0 9/30/98 B 3.6.4-1 Revision 0 9/30/98 B 3.6.4-2 Revision 1 2/26/01 B 3.6.4-3 Revision 0 9/30/98 B 3.6.4-4 Revision 0 9/30/98 B 3.6.5-1 Revision 0 9/30/98 B 3.6.5-2 Revision 1 4/26/00 B 3.6.5-3 Revision 1 4/26/00 B 3.6.5-4 Revision 0 9/30/98 B 3.6.6-1 Revision 0 9/30/98 B 3.6.6-2 Revision 0 9/30/98 B 3.6.6-3 Revision 0 9/30/98 B 3.6.6-4 Revision 0 9/30/98 B 3.6.6-5 Revision 0 9/30/98 B 3.6.6-6 Revision 2 4/26/00 B 3.6.6-7 Revision 1 4/26/00 B 3.6.7-1 Revision 0 9/30/98 B 3.6.7-2 Revision 0 9/30/98 B 3.6.7-3 Revision 1 4/29/04 B 3.6.7-4 Revision 1 4/29/04 B 3.6.7-5 Revision 0 9/30/98 B 3.6.8-1 Revision 1 4/26/00 B 3.6.8-2 Revision 0 9/30/98 B 3.6.8-3 Revision 2 3/01/05 B 3.6.8-4 Revision 2 4/29/04 B 3.6.8-5 Revision 0 9/30/98 B 3.6.9-1 Revision 1 5/05/00 B 3.6.9-2 Revision 2 3/01/05 B .3.6.9-3 Revision 1 5/05/00 B 3.6.9-4 Revision 1 5/05/00 Catawba Units 1 and 2Pae23/90 Page 25 3/19/08

Page Number Amendment Revision Date B 3.6.9-5 Revision 3 -5/10/05 B 3.6.9-6 Revision 2 2/26/01 B 3.6.10-1 Revision 1 9/30/05 B 3.6.10-2 Revision 1 9/30/05 B 3.6.10-3 Revision 1 9/30/05 B 3.6.10-4 Revision 1 9/30/05 B 3.6.10-5 Revision 1 9/30/05 B 3.6.10-6 Revision 1 9/30/05 B 3.6.11 -1 Revision 0 9/30/98 B 3.6.11-2 Revision 0 9/30/98 B 3.6.11-3 Revision. 0 9/30/98 B 3.6.11-4 Revision 1 2/26/99 B 3.6.11-5 Revision 2 2/26/99 B 3.6.12-1 Revision 3 5/10/05 B 3.6.12-2 Revision 2 5/10/05 B 3.6.12-3 Revision 3 5/10/05 B 3.6.12-4 Revision 3 5/10/05 B 3.6.12-5 Revision 2 5/10/05 B 3.6.12-6 Revision 4 5/10/05 B 3.6.12-7 Revision 3 5/10/05 B 3.6.12-8 Revision 2 5/10/05 B 3.6.12-9 Revision 3 5/10/05 B 3.6.12-10 Revision 2 5/10/05 B 3.6.12-11 Revision 1 5/10/05 B 3.6.13-1 Revision 0 9/30/98 B 3.6.13-2 Revision 0 9/30/98 B 3.6.13-3 Revision 0 9/30/98 B 3.6.13-4 Revision 0 9/30/98 B 3.6.13-5 Revision 0 9/30/98 B 3.6.13-6 Revision 0 9/30/98 B 3.6.13-7 Revision 0 9/30/98 B 3.6.13-8 Revision 2 12/4/06 Catawba Units 1 and 2Pae23/90 Page 26 3/19/08

Page Number Amendment Revision Date B 3.6.13-9 Revision 1 12/4/06 B 3.6.14-1 Revision 0 9/30/98 B 3.6.14-2 Revision 0 9/30/98 B 3.6.14-3 Revision 0 9/30/98 B 3.6.14-4 Revision 0 9/30/98 B 3.6.14-5 Revision 0 9/30/98 B 3.6.14-6 Revision 0 9/30/98 B 3.6.15-1 Revision 0 9/30/98 B 3.6.15-2 Revision 0 9/30/98 B 3.6.15-3 Revision 0 9/30/98 B 3.6.15-4 Revision 0 9/30/98 B 3.6.16-1 Revision 1 4/09/99 B 3.6.16-2 Revision 2 9/30/05 B 3.6.16-3 Revision 2 9/30/05 B 3.6.16-4 Revision 0 9/30/05 B 3.6.17-1 Revision 1 3/13/08 B 3.6.17-2 Revision 0 9/30/98 B 3.6.17-3 Revision 0 9/30/98 B 3.6.17-4 Revision 0 9/30/98 B 3.6.17-5 Revision 1 3/13/08 B 3.7.1-1 Revision 0 9/30/98 B 3.7.1-2 Revision 0 9/30/98 B 3.7.1-3 Revision 0 9/30/98 B 3.7.1-4 Revision 0 9/30/98 B 3.7.1-5 Revision 0 9/30/98 B 3.7.2-1 Revision 0 9/30/98 B 3.7.2-2 Revision 0 9/30/98 B 3.7.2-3 Revision 1 3/13/08 B 3.7.2-4 Revision 0 9/30/98 B 3.7.2-5 Revision 1 3/13/08 B 3.7.3-1 Revision O 9/30/98 B 3.7.3-2 Revision 0 9/30/98 Catawba Units 1 and 2 Page 27 3/19/08

Page Number Amendment Revision Date B 3.7.3-3 Revision 0 9/30/98 B 3.7.3-4 Revision 0 9/30/98 B 3.7.3-5 Revision 0 9/30/98 B 3.7.3-6 Revision 0 9/30/98 B 3.7.4-1 Revision 0 9/30/98 B 3.7.4-2 Revision 0 9/30/98 B 3.7.4-3 Revision 1 4/29/04 B 3.7.4-4 Revision 0 9/30/98 B 3.7.5-1 Revision 0 9/30/98 B 3.7.5-2 Revision 0 9/30/98 B 3.7.5-3 Revision 0 9/30/98 B 3.7.5-4 Revision 2 4/29/04 B 3.7.5-5 Revision 1 4/29/04 B 3.7.5-6 Revision 1 4/29/04 B 3.7.5-7 Revision .1 4/29/04 B 3.7.5-8 Revision 1 4/29/04 B 3.7.5-9 Revision 1 4/29/04 B 3.7.6-1 Revision 2 5/10/05 B 3.7.6-2 Revision 1 5/10/05 B 3.7.6-3 Revision 2 3/19/08 B 3.7.6-4 Revision 0 9/30/98 B 3.7.7-1 Revision 1 5/10/05 B 3.7.7-2 Revision 0 9/30/98 B 3.7.7-3 Revision 0 9/30/98 B 3.7.7-4 Revision 0 9/30/98 B 3.7.7-5 Revision 0 -9/30/98 B 3.7.8-1 Revision 0 9/30/98 B 3.7.8 Revision 0 9/30/98 B 3.7.8-3 Revision 1 2/26/99 B 3.7.8-4 Revision.1 2/26/99 B 3.7.8-5 Revision 1 2/26/99 B 3.7.9-1 Revision 2 9/25/06 Catawba Units 1 and 2Pae23/90 Page 28 3/19/08

Page Number Amendment Revision Date B 3.79-2 Revision 2 9/25/06 B 3.7.9-3 Revision 1 9/25/06 B 3.7.9-4 Revision 1 9/25/06 B 3.7.10-1 Revision 2 9/30/05 B 3.7.10-2 Revision 4 3/13/08 B 3.7.10-3 Revision 6 3/130/08 B 3.7.10-4 Revision 4 9/30/05 B 3.7.10-5 Revision 5 9/30/05 B 3.7.10-6 Revision 3 9/30/05 B 3.7.10-7 Revision 3 B 3.7.11 -1 Revision 0 9/30/98 B 3.7.11-2 Revision 1 4/23/02 B 3.7.11-3 Revision 1 4/23/02 B 3.7.11-4 Revision 1 4/23/02 B 3.7.12-1 Revision 2 9/30/05 B 3.7.12-2 Revision 2 9/30/05 B 3.7.12-3 Revision 2 9/30/05 B 3.7.12-4 Revision 2 9/30/05 B 3.7.12-5 Revision 2 9/30/05 B 3.7.12-6 Revision 1 9/30/05 B 3.7.12-7 Revision 0 9/30/05 B 3.7.13-1 Revision 3 9/30/05 B 3.7.13-2 Revision 3 9/30/05 B 3.7.13-3 Revision 2 9/30/05 B 3.7.13-4 Revision 2 9/30/05 B 3.7.13-5 Revision 2 9/30/05 B 3.7.14-1 Revision 1 3/13/08 B 3.7.14-2 Revision 0 9/30/98 B 3.7.14-3 Revision 1 3/13/08 B 3.7.15-1 Revision 1 9/27/06 B 3.7.15-2 Revision 1 9/27/06 B 3.7.15-3 Revision 1 9/27/06 Catawba Units 1 and 2Pae23/90 Page 29 3/19/08

Page Number Amendment Revision Date B 3.7.15-4 Revision 0 9/27/06 B 3.7.16-1 Revision 2 9/27/06 B 3.7.16-2 Revision 2 9/27/06 B 3.7.16-3 Revision 2 9/27/06 B 3.7.16-4 Revision 0 9/27/06 B 3.7.17-1 Revision 1 3/13/08 B 3.7.17-2 Revision 1 3/13/08 B 3.7.17-3 Revision 1 3/13/08 B 3.8.1 -1 Revision 1 2/26/01 B 3.8.1-2 Revision 0 9/30/98 B 3.8.1-3 Revision 1 11/5/07 B 3.8.1-4 Revision 2 11/5/07 B 3.8.1-5 Revision 2 5/10/05 B 3.8.1-6 Revision 2 5/10/05 B 3-.8.1-7 Revision 2 5/10/05 B 3.8.1-8 Revision 1 5/10/05 B 3.8.1-9 Revision 1 5/10/05 B 3.8.1 -10 Revision 1 5/10/05 B 3.8.1 -11 Revision 0 9/30/98 B 3.8.1-12 Revision 0 9/30/98 B 3.8.1-13 Revision 0 9/30/98 B,3.8.1-14 Revision 0 9/30/98 B 3.8.1-15 Revision 0 9/30/98 B 3.8.1-16 Revision 0 9/30/98 B 3.8.1-17 Revision 0 9/30/98 B 3.8.1-18 Revision 1 11/5/07 B 3.8.1-19 Revision 1 11/5/07 B 3.8.1-20 Revision 0 9/30/98 B 3.8.1-21 Revision 0 9/30/98 B 3.8.1-22 Revision 2 6/25/07 B 3.8.1-23 'Revision 1 3/16/00 B 3.8.1-24 Revision 0 9/30/98 Catawba Units 1 and 2Pae33/90 ,Page 30 3/19/08

Page Number Amendment Revision Date B 3.8.1-25 Revision 0 9/30/98 B 3.8.1-26 Revision 0 9/30/98 B 3.8.1-27 Revision 2 6/25/07 B 3.8.2-1 Revision 0 9/30/98 B 3.8.2-2 Revision 0 9/30/98 B 3.8.2-3 Revision 0 9/30/98 B 3.8.2-4 Revision 1 5/10/05 B 3.8.2-5 Revision 2 5/10/05 B 3.8.2-6 Revision 1 5/10/05 B 3.8.3-1 Revision 1 1/15/99 B 3.8.3-2 Revision 0 9/30/98 B 3.8.3-3 Revision 1 1/15/99 B 3.8.3-4 Revision 0 9/30/98 B 3.8.3-5 Revision 1 1/15/99 B 3.8.3-6 Revision 1 7/10/03 B 3.8.3-7 Revision 1 7/10/03 B 3.8.3-8 Revision 2 5/10/05 B 3.8.4-1 Revision 0 9/30/98 B 3.8.4-2 Revision 1 2/26/99 B 3.8.4-3 Revision 0 9/30/98 B 3.8.4-4 Revision 1 4/27/99 B 3.8.4-5 Revision 3 4/27/05 B 3.8.4-6 Revision 4 4/27/05 B 3.8.4-7 Revision 7 4/27/05 B 3.8.4-8 Revision 6 10/06/05 B 3.8.4-9 Revision 6 B 3.8.4-10 Revision 1 3/29/05 B 3.8.5-1 Revision 0 9/30/98 B 3.8.5-2 Revision 2 7/29/03 B 3.8.5-3 Revision 1 7/29/03 B 3.8.6-1 Revision 2 4/27/05 B 3.8.6-2 Revision 1 4/27/05 Catawba Units 1 and 2Pge33/90 Page 31 3/19/08

Page Number Amendment Revision Date B 3.8.6-3 Revision 2 4/27/05 B 3.8.6-4 Revision 2 4/27/05 B 3.8.6-5 Revision 1 4/27/05 B 3.8.6-6 Revision 1 4/27/05 B 3.8.6-7 Revision 1 4/27/05 B 3.8.7-1 Revision 0 9/30/98 B 3.8.7-2 Revision 1 3/15/04 B 3.8.7-3 Revision 2 3/15/04 B 3.8.7-4 Revision 0 3/15/04 B 3.8.8-1 Revision 0 9/30/98 B 3.8.8-2 Revision 2 10/10/06 B 3.8.8-3 Revision 2 7/29/03 B 3.8.8-4 Revision 0 7/29/03 B 3.8.9-1 Revision 0 9/30/98 B 3.8.9-2 Revision 0 9/30/98 B 3.8.9-3 Revision 0 9/30/98 B 3.8.9-4 Revision 0 9/30/98 B 3.8.9-5 Revision 0 9/30/98 B 3.8.9-6 Revision 0 9/30/98 B 3.8.9-7 Revision 0 9/30/98 B 3.8.9-8 Revision 0 9/30/98 B 3.8.9-9 Revision 1 2/26/99 B 3.8.9-10 Revision 1 2/26/99 B 3.8.10-1 Revision 0 9/30/98 B 3.8.10-2 Revision 0 9/30/98 B 3.8.10-3 Revision 2 7/29/03 B 3.8.10-4 Revision 1 7/29/03 B 3.9.1 -1 Revision 1 9/1/05 B 3.9.1-2 Revision 2 9/1/05 B 3.9.1-3 Revision 3 9/1/05 B 3.9.1-4 Revision 0 9/1/05 B 3.9.2-1 Revision 2 6/21/04 Catawba Units 1 and 2Pae33/90 Page 32 3/19/08

Page Number Amendment Revision Date B 3.9.2-2 Revision 2 6/21/04 B 3.9.2-3 Revision 2 6/21/04 B 3.9.3-1 Revision 3 3/13/08 B 3.9.3-2 Revision 1 4/23/02 B 3.9.3-3 Revision 3 3/13/08 B 3.9.3-4 Revision 3 3/13/08 B 3.9.3-5 Revision 2 9/30/05 B 3.9.4-1 Revision 0 9/30/98 B 3.9.4-2 Revision 3 11/24/04 B 3.9.4-3 Revision 2 11/24/04 B 3.9.4-4 Revision 2 11/24/04 B 3.9.5-1 Revision 0 9/30/98 B 3.9.5-2 Revision 2 7/29/03 B 3.9.5-3 Revision 1 7/29/03 B 3.9.5-4 Revision 1 7/29/03 B 3.9.6-1 Revision 1 3/13/08 B 3.9.6-2 Revision 0 9/30/98 B 3.9.6-3 Revision 1 3/13/08 B 3.9.7-1 Revision 0 6/21/04 B 3.9.7-2 Revision 0 6/21/04 B 3.9.7-3 Revision 0 6/21/04 Catawba Units 1 and 2 Page 33 3/19/08

SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) safety injection is blocked, administrative controls on boron concentration are required to prevent a post-trip return-to-power.

Another limiting event for the SDM requirements is the inadvertent boron dilution event. The postulated reduction in the moderator boron concentration causes an insertion of positive reactivity. This positive reactivity insertion, if not terminated by operiator or automatic action, would eventually cause a return to criticality and potentially to power operation. The expected method of stopping a dilution event initiated from a shutdown condition is the Boron Dilution Mitigation System (BDMS). However, in the event the BDMS is not OPERABLE, the safety analyses rely on operator action. MODE 5, because of its combination of a lower RCS liquid volume, which results in a faster dilution, and the lower SDM limit, is the limiting operational mode for this accident. The licensing basis acceptance criterion is that the operator has at least fifteen minutes available, from the time he is alerted to the occurrence of a dilution event (by the High Flux at Shutdown alarm) to the time of criticality, to terminate the dilution. The most limiting boron dilution is one initiated at the beginning of core life since the high RCS boron concentrations at that burnup are more easily reduced, resulting in a larger concentration change per unit of time and therefore, a faster positive reactivity insertion. In order to meet the fixed operator action time criterion for delaying criticality, this faster dilution may be compensated for by a larger SDM (Ref. 4).

In addition to the limiting MSLB transient, the SDM requirement must also protect against:

a. An uncontrolled rod withdrawal from subcritical or low power condition; and

b. Rod ejection.

Each of these events is discussed below.

Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power.

level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures dnd pressure.

The ejection of a rod also produces a time dependent redistribution of core power. SDM satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3). Even Catawba Units 1 and 2 B 3.1.1-3 Revision No. 1

SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the. unit is operating within the bounds of accident analysis assumptions.

Transients which are made less severe by the rapid insertion of control rod negative reactivity are also affected by the magnitude of the SDM limit. This is because the safety analyses assume a change in the rate of insertion of this negative reactivity when the SDM limit is reached. While the SDM is less than .the limit value, the negative reactivity from the control rods is assumed to be inserted as quickly as the rod worth vs.

time curves shown in Reference 5. When the SDM limit value is reached, the rate of negative reactivity insertion is decreased so that it is only fast enough to compensate for any positive reactivity insertion, e.g., from the cooling of the fuel and moderator (which normally have negative temperature coefficients). This methodology is conservative in that it does not take credit in the safety analyses, even temporarily, for a SDM greater than the limit value.

LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

The MSLB (Ref. 2) and the boron dilution (Ref. 4) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 50.67 limits (Ref. 5). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.

APPLICABILITY In MODE 2 with ke,, < 1.0 and in MODES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2 with keff ->1.0, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits,"

and LCO 3.1.6.

Catawba Units 1 and 2 B 3.1.1-4 Revision No. 2

SDM B 3.1.1 BASES ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly.

A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.

In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the boric acid storage tank, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.

In determining the boration flow rate, the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle when the boron concentration may approach or exceed 2000 ppm. Using its normal makeup path, the Chemical and Volume Control System (CVCS) is capable of inserting negative reactivity at a. rate of approximately 30 pcm/min when the RCS boron concentration is 1000 ppm and approximately 35 pcm/min when the RCS boron concentration is 100 ppm. If the emergency boration path is used, the CVCS is capable of inserting negativereactivity at the rate of 65 pcm/min when the RCS boron concentration is 1000 ppm and 75 pcm/min when the RCS boron concentration is 100 ppm. Therefore, if SDM had to be increased by 1%

Ak/k or 1000 pcm, normal makeup path at 1000 ppm could restore SDM in approximately 33 minutes. At 100 ppm, SDM could be restored in approximately 29 minutes. In the emergency boration mode at 1000 ppm, the 1% Ak/k could be restored in approximately 15 minutes. With RCS boron concentration at 100 ppm, SDM could be increased by 1000 pcm in approximately 13 minutes using emergency boration. These boration parameters represent typical values and are provided for the purpose of offering a specific example.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2 with keff - 1.0, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.

Catawba Units 1 and. 2 B 3.1.1-5 Revision No. 1

SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

In MODE 2 with keof < 1.0 and MODES 3, 4, and 5, SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:

a. RCS boron concentration;

b. Control bank position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the generally slow change in required boron concentration and the low probability of an accident occurring without the required SDM. This allows time for the operator to collect the required data, which includes performing a boron concentration analysis, and complete the calculation.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. UFSAR, Section 15.1.5.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. UFSAR, Section 15.4.6.

5. 10 CFR 50.67.

Catawba Units 1 and 2 B 3.1.1-6 Revision No. 2

RTS Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Trip System (RTS) Instrumentation BASES BACKGROUND The RTS initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during anticipated operational occurrences (AOOs) and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents.

The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RTS, as well as specifying LCOs on other reactor system parameters and equipment performance.

The LSSS, defined in this specification as the Allowable Value, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During AOOs, which are those events expected to occur one or more times during the unit life, the acceptable limits are:

1. The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to prevent departure from nucleate boiling (DNB);

2. Fuel centerline melt shall not occur; and

3. The RCS pressure SL of 2735 psig shall not be exceeded.

Operation within the SLs of Specification 2.0, "Safety Limits (SLs)," also maintains the above values and assures that offsite dose will be within the 10 CFR 20 and 10 CFR 50.67 criteria during AOOs.

Accidents are events that are analyzed even though they are not expected to occur during the unit life. The acceptable limit during accidents is that offsite dose shall be maintained within an acceptable fraction of 10 CFR 50.67 limits. Different accident categories are allowed a different fraction of these limits, based on probability of occurrence.

Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

Catawba Units 1 and 2 B 3.3.1 -1 Revision No. 2

RTS Instrumentation 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One channel inoperable. ------------------- NOTE -------------------

The inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing and setpoint adjustment of other channels.

D.1.1 Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND D.1.2 Reduce THERMAL 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months POWER to < 75% RTP.

OR D.2.1 Place channel in trip. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND D.2.2 --------- NOTE -------

Only required to be performed when the Power Range Neutron Flux input to QPTR is inoperable.

Perform SR 3.2.4.2. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OR D.3 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued)

Catawba Units 1 and 2 3.3-1-2 Amendment Nos. 173/165

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) time could be affected is replacing the sensing assembly of a.transmitter.

As appropriate, each channel's response must be verified every 18 months on a STAGGERED TEST BASIS. Testing of the final actuation devices is included in the testing. Testing of the RTS RTDs is performed on an 18 month frequency. Response times cannot be determined during unit operation because equipment operation is required to measure response times. Experience has shown that these components usually pass this surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. The response time of the neutron flux signal portion of the channel shall be measured from detector output or input of the first electronic component in the channel.

REFERENCES 1. UFSAR, Chapter 7.

2. UFSAR, Chapter 6.

3. UFSAR, Chapter 15.

4. IEEE-279-1971.

5. 10 CFR 50.49.

6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

7. WCAP-10271-P-A, Supplement 2, Rev. 1, June "1990.

8. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" Sep., 1995.

9. WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" Oct., 1998.

10.10 CFR 50.67.

Catawba Units 1 and 2 B 3.3.1-51 Revision No. 2

Containment Air Release and Addition Isolation Instrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Containment Air Release and Addition Isolation Instrumentation BASES BACKGROUND Containment air release and addition isolation instrumentation closes the containment isolation valves in the Containment Air Release and Addition System. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident.

Containment air release and addition isolation initiates on an automatic safety injection (SI) signal through the Containment Isolation-Phase A Function, or by manual actuation of Phase A Isolation. The Bases for LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation," discuss these modes of initiation.

Each of the containment air release and addition penetrations has inner and outer containment isolation valves. A safety injection initiates containment isolation, which closes both inner and outer containment isolation valves. The Containment Air Release and Addition System is described in the Bases for LCO 3.6.3, "Containment Isolation Valves."

APPLICABLE The safety analyses assume that the containment remains SAFETY ANALYSES intact with penetrations unnecessary for core cooling isolated early in the event, within approximately 60 seconds. The Containment Air Release and Addition System isolation valves may be used in MODES 1-4 and their rapid isolation is assumed. Containment isolation ensures meeting the containment leakage rate assumptions of the safety analyses, and -

ensures that the calculated accidental offsite radiological doses are below 10 CFR 50.67 (Ref. 1) limits.

The containment air release and addition isolation instrumentation satisfies Criterion 3 of 10 CFR 50.36 (Ref. 2)..

Catawba Units 1 and 2 B 3.3.6-1 Revision No. 2

Containment Air Release and Addition Isolation Instrumentation B 3.3.6 BASES LCO The LCO-requirements ensure that the instrumentation necessary to initiate Containment Air Release and Addition Isolation, listed in Table 3.3.6-1., is OPERABLE.

1. Manual Initiation The LCO requires two trains OPERABLE. The operator can initiate containment isolation at any iime by using either of two switches (manual Phase A actuation or manual spray actuation) in the control room. Either switch actuates its associated train. This action will cause actuation of all components in the same manner as any of the automatic actuation signals.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each train consists of one push button and the interconnecting wiring to the actuation logic cabinet.

2. Automatic Actuation Logic and Actuation Relays The: LCO requires two trains of Automatic Actuation Logic and Actuation RelaysOPERABLE to ensure that no single random failure can prevent automatic actuation.

Automatic Actuation Logic and Actuation Relays consist of the same features and operate in the same manner as described for ESFAS Function 1.b, SI, and ESFAS Function 3.a, Containment Phase A Isolation. The applicable MODES and specified conditions for the containment air release. and addition isolation portion of these Functions are different and less restrictive than -

those for their Phase A isolation and SI roles. If one or more of the SI or Phase A isolation Functions becomes inoperable in such a manner that only the containment air release and addition isolation Function is affected, the Conditions applicable to their SI and Phase A isolation Functions need not be entered. The less restrictive Actions specified for inoperability of the containment air release and addition isolation Functions specifysufficient compensatory measures for this case.

Catawba Units 1 and 2 B 3.3.6-2 Revision No. 1

Containment Air Release-and Addition Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued)

For slave relays or any auxiliary relays in the circuit that are of the type Westinghouse AR or Potter & Brumfield MDR, the SLAVE RELAY TEST is performed every 18 months. This test frequency is based on the relay reliability assessments presented in References 3, 4, and 5. These reliability assessments are relay specific and apply only to the Westinghouse AR and Potter & Brumfield MDR type relays. SSPS slave relays or any auxiliary relays not addressed by Reference 3 do not qualify for extended surveillance intervals and will continue to be tested at a 92 day Frequency.

SR 3.3.6.4 SR 3.3.6.4 is the performance of a TADOT. This test is a check of the Manual Actuation Functions and is performed every 18 months. Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.).

The test also includes trip devices that provide actuation signals directly to the SSPS, bypassing the analog process control equipment. The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

The Frequency is based on the known reliability of the Function and the redundancy available, and has been shown to be acceptable through operating experience.

REFERENCES 1. 10 CFR 50.67.

2. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

3. WCAP-13900, "Extension of Slave Relay Surveillance Test Intervals," April 1994.

4. WCAP-13877 Revision 2-P-A, "Reliability Assessment of Westinghouse Type AR Relays Used as SSPS Slave Relays,"

August 2000.

5. WCAP-13878-P-A Revision 2, "Reliability Assessment of Potter &

Brumfield MDR Series Relays," August 2000.

Catawba Units 1 and 2 B 3.3.6-5 Revision No. 3

RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY ANALYSES address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event.

Catawba Units 1 and 2 B 3.4.13-1 Revision No. 0

RCS Operational LEAKAGE B 3.4.13 BASES APPLICABLE SAFETY ANALYSES (continued)

.The safety analysis (Ref. 3) for an event resulting in steam discharge to the atmosphereassumes that primary to secondary LEAKAGE from each steam generator (SG) is 150 gallons per day. Any event in which the reactor coolant system will continue to leak water inventory to the secondary side, and in which there will be a postulated source term associated with the accident, utilizes this leakage value as an input in the analysis. These accidents include the rod ejection accident, locked rotor accident, main steam line break, steam generator tube rupture and uncontrolled rod withdrawal accident. The rod ejection accident, locked rotor accident and uncontrolled rod withdrawal accident yield a source term due to postulated fuel failure as a result of the accident. The main steam line break and the steam generator tube rupture yield a source term due to perforations in fuel pins causing an iodine spike. Primary to secondary side leakage may escape the secondary side due to flashing or atormization of the coolant, or it may mix with the secondary side SG water inventory and be released due to steaming of the SGs. The rod ejection accident is limiting compared to the remainder of the accidents with respect to dose results. The dose results for each of the accidents delineated above are below the 10 CFR 50.67 limits (Ref. 9) and the limits in Regulatory Guide 1.183 (Ref. 10) for these accidents.

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36 (Ref. 4).

LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE.

Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment Catawba Units 1 and 2 B *3.4.13-2 Revision No. 2

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively Identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and Identified LEAKAGE are determined by performance of an RCS water inventory balance. For this SR, the volumetric calculation of unidentified LEAKAGE and identified LEAKAGE is based on a density at room temperature of 77 degrees F.

The Surveillance is modified by two Notes. The RCS water inventory balance must be performed with the reactor at steady state operating conditions and near operating pressure. Therefore, Note 1 indicates that this SR is not required to be completed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established.

Steady state operation is required to perform a proper inventory balance; calculations during maneuvering are not useful and Note 1 requires the Surveillance to be met when steady state is established. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

Note 2 States that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day or lower cannot be measured accurately by an RCS water inventory balance.

An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents and reduction of potential consequences. A Note under the Frequency column states that this SR is only required to be performed during steady state operation.

Catawba Units I and 2 B 3.4.13-5 Revision No. 3

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.18, "Steam Generator (SG) Tube Integrity," should be evaluated. The 150 gallons per day limit is based on measurements taken at room temperature, with a correction factor applied to account for the fact that current safety analyses take the primary to secondary leak rate at reactor coolant conditions, rather than at room temperature.

The Surveillance is modified by a Note which states that this SR is not required to be completed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of steady state operation near operating pressure have been established. During normal operation the primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents and reduction of potential consequences. A Note under the Frequency column states that this SR is only required to be performed during steady state operation.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 30.

2. Regulatory Guide 1.45, May 1973.

3. UFSAR, Section 15.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. EPRI TR-104788-R2, "PWR Primary-to-Secondary Leak Guidelines," Revision 2.

6. NEI 97-06, "Steam Generator Program Guidelines."

7. UFSAR, Section 18, Table 18-1.

8. Catawba License Renewal Commitments, CNS-1274.00-00-0016, Section 4.27.

9. 10 CFR 50.67.

Catawba Units 1 and 2 B 3.4.13-6 Revision No. 4

RCS Operational LEAKAGE B 3.4.13 BASES REFERENCES (continued)

10. Regulatory Guide 1.183, July 2000.

Catawba Units 1 and 2 B 3.4.13-7 Revision No. 0

RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES BACKGROUND The maximum total effective dose equivalent an individual at the site boundary can receive for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during anaccident is specified in 10 CFR 50.67 (Ref. 1). The limits on specific activity ensure that the doses are held below 10 CFR 50.67 and Regulatory Guide 1.183 limits during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose at the site boundary to within the acceptance criteria of 10 CFR 50.67 and Regulatory Guide 1.183. The limits in the LCO are standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site locations.

The parametric evaluations showed the potential offsite dose levels for a SGTR accident were within the acceptance criteria of 10 CFR 50.67 and Regulatory Guide 1.183. Each evaluation assumes a broad range of site applicable atmospheric dispersion factors in a parametric evaluation.

APPLICABLE The LCO limits on the specific activity of the reactor coolant ensures that SAFETY ANALYSES the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed the 10 CFR 50.67 and Regulatory Guide 1.183 acceptance criteria. The SGTR safety analysis (Ref. 2) assumes the specific activity of the reactor coolant at the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 150 gpd per SG. The safety analysis assumes the specific activity of the secondary coolant at its limit of 0.1 pCi/gm DOSE EQUIVALENT 1-131 from LCO 3.7.17, "Secondary Specific Activity."

The analysis for the SGTR accident establishes the acceptance limits for RCS specific activity. Reference to this analysis is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

Catawba Units 1 and 2 B 3.4.16-1 Revision No. 1

RCS Specific Activity B 3.4.16 BASES APPLICABLE SAFETY ANALYSES (continued)

The analysis is for two cases of reactor coolant specific activity. One case assumes specific activity at 1.0 pCi/gm DOSE EQUIVALENT 1-131 with a concurrent large iodine spike that increases the 1-131 activity in the .

reactor coolant by a factor of about 50 immediately after the accident.

The second case assumes the initial reactor coolant iodine activity at 60.0 pCi/gm DOSE EQUIVALENT 1-131 due to a pre-accidentiodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coolant assumes 1% failed fuel, which closely equals the LCO limit of 100/E pCi/gm for gross specific activity.

The analysis also assumes a loss of offsite power at the same time as the SGTR event. The SGTR causes a reduction in reactor coolant inventory.

The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature AT signal if the leakage continues for a long enough time, although a manual trip is also credited after a conservatively long delay.

The coincident loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves.

The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends.

The safety analysis shows the radiological consequences of an SGTR accident are within dose-guideline limits of References 1 and 4.

Operation with iodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed the limits shown in Figure 3.4.16-1, in the applicable specification, for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The safety analysis has concurrent and pre-accident iodine spiking levels up to 60.0 RCi/gm DOSE EQUIVALENT 1-131.

The remainder of the above limit permissible iodine levels shown in Figure 3.4.16-1 are acceptable because of the low probability of a SGTR accident occurring during the established 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months time limit. The occurrence of an SGTR accident at these permissible levels could increase the site boundary dose levels, but still be within 10 CFR 50.67 dose guideline limits.

The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.

RCS specific activity satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3).

Catawba Units 1 and 2 B 3.4.16-2 Revision No. 1

RCS Specific Activity B .3.4.16 BASES LCO The specific iodine activity is limited to 1.0 !tCi/gm DOSE EQUIVALENT 1-131, and the gross specific activity in the reactor coolant is limited to the number of [tCi/gm equal to 100 divided by E (average disintegration energy of the sum of the average beta and gamma energies of the coolant nuclides). The limit on DOSE EQUIVALENT 1-131 and the limit on gross specific activity ensure the maximum 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months total effective dose equivalent to an individual at the site boundary during the Design Basis Accident (DBA) will be within the limits of 10 CFR 50.67 and Regulatory Guide 1.183.

The SGTR accident analysis (Ref. 2) shows that the maximum 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months site boundary dose levels are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of an SGTR, lead to site boundary doses that exceed the 10 CFR 50.67 dose guideline limits.

APPLICABILITY In MODES 1 and 2, and in MODE 3 with RCS average temperature

>_500 0 F, operation within the LCO limits for DOSE EQUIVALENT 1-131 and gross specific activity are necessary to contain the potential consequences of an SGTR to within the acceptable site boundary dose values.

For operation in MODE 3 with RCS average temperature < 5000 F, and in

-MODES 4 and 5, the release of radioactivity in the event of a SGTR is unlikely since the saturation pressure of the reactor coolant is below the lift pressure settings of the main steam safety valves.

ACTIONS A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months must be taken to demonstrate that the limits of Figure 3.4.16-1 are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.

The DOSE EQUIVALENT 1-131 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is required, if the limit

-violation resulted from normal iodine spiking.

A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS.

This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore Catawba Units 1 and 2 B 3.4.16-3 Revision No. 2

RCS Specific Activity B 3.4.16 BASES ACTIONS (continued) transient specific activity excursions while the plant remains at, or proceeds to power operation.

B.1 With the gross specific activity in excess of the allowed limit, the unit must be placed in a MODE in which the requirement does not apply.

The change within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to MODE 3 and RCS average temperature

< 500OF lowers the saturation pressure of the reactor coolant below the setpoints of the main steam safety valves and prevents venting the SG to the environment in an SGTR event. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.

C.1 If a Required Action and the associated Completion Time of Condition A is not met or if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.16-1, the reactor must be brought to MODE 3 with RCS average temperature < 500OF within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 below 500OF from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at least once every 7 days. A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 10 minutes and all radioiodines.

The total specific activity shall be the sum of the beta-gamma activity in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radionuclides. This Surveillance provides an indication of any increase in gross specific activity.

Catawba Units 1 and 2 B 3.4.16-4 Revision No. 2

RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued)

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.

The Surveillance is applicable in MODES 1 and 2, and in MODE 3 with Tavg at least 500 0 F. The 7 day Frequency considers the unlikelihood of a gross fuel failure during the time.

SR 3.4.16.2 This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering gross activity is monitored every 7 days. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change Ž 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results. If the power excursion is one continuous process spanning over several hours, there is no need to sample every hour, only 2 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after the last major power change of

> 15% RTP, since this sample will encompass the maximum potential for additional iodine release to have occurred.

SR 3.4.16.3 A radiochemical analysis for E determination is required every 184 days (6 months) with the plant operating in MODE 1 equilibrium conditions.

The E- determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for f is a measurement of the average energies per disintegration for isotopes with half lives longer than 10 minutes, excluding iodines. The Frequency of 184 days recognizes E does not change rapidly.

This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

Catawba Units 1 and 2 B 3.4.16-5 Revision No. 2

RCS Specific Activity B 3.4.16 BASES REFERENCES 1. 10 CFR 50.67.

2. UFSAR, Section 15.6.3.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. Regulatory Guide 1.183, July 2000.

Catawba Units 1 and 2 B 3.4.16-6 Revision No. 1

SG Tube Integrity B 3.4.18 BASES APPLICABLE SAFETY ANALYSES (continued) and the main steam code safety valves until such time as the closure of these valves can be credited.

For other design basis accidents such as main steam line break, rod ejection accident, reactor coolant pump locked rotor accident, and uncontrolled rod withdrawal accident, the tubes are assumed to retain their structural integrity (i.e;, they are assumed not to rupture). The LEAKAGE is assumed to be initially at the limit given in LCO 3.4.13.

The three SG performance criteria and the limits included in LCO 3.4.16, "RCS Specific Activity," for DOSE EQUIVALENT 1-131 in primary coolant, and in LCO 3.7.17, "Secondary Specific Activity,"

for DOSE EQUIVALENT 1-131 in secondary coolant, ensure the plant is operated within its analyzed condition. The dose consequences resulting from the most limiting design basis accident are within the limits defined in GDC 19 (Ref. 2), 10 CFR 50.67 (Ref. 4), and Regulatory Guide 1.183 (Ref. 3).

SG Tube Integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During a SG inspeclion, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall and any repairs made to it, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.9, "Steam Generator (SG) Program," and describe acceptable SG tube performance. The Steam GeneratorProgram also provides the evaluation process for determining conformance with the SG performance criteria.

Catawba Units 1 and 2 B 3.4.18-3 Revision No. 1

SG Tube Integrity B 3.4.18 BASES LCO (continued)

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 5) and Draft Regulatory Guide 1.121 (Ref.

6)., Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." Significant is defined as, "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of structural integrity performance criterion causes a lower structural limit or limiting burst/collapse condition to be established."

The accident induced leakage performance criterion ensures that the primary to secondary LEAKAGE caused by a design basis accident, other than a SG tube rupture, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 150 gallons per day through each SG for a total of 600 gallons per day through all SGs. The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons, per day.

This limit is based on the assumption that a single crack leaking Catawba Units 1 and 2 B 3.4.18-4 Revision No. 0

SG Tube Integrity B 3.4.18 BASES SURVEILLANCE REQUIREMENTS (continued)

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.18.1. The Frequency is determined in part by the operational assessment and other limits in the Steam Generator Examination Guidelines (Ref. 7). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.5.9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 3.4.18.2 During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 5.5.9 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performa nce criteria will continue to be met until the next inspection of the subject tube(s). Ref. 1 and Ref. 7 provide guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following a SG tube inspection ensures that the Surveillance has been completed and all tubes satisfying the repair criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.

Catawba Units 1 and 2 B83.4.18-7 Revision No'. 0

SG Tube Integrity B 3.4.18 BASES REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50 Appendix A, GDC 19.

3. Regulatory Guide 1.183, July 2000. I

4. 10 CFR 50.67.

5.. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

6. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

7. EPRI TR-107569, "Pressurized Water Reactor Steam Generator Examination Guidelines."

Catawba Units 1 and 2 B 3.4.18-8 Revision No. I

CVIWS B 3.6.17 B 3.6 CONTAINMENT SYSTEMS B 3.6.17 Containment Valve Injection Water System (CVIWS)

BASES BACKGROUND The CVIWS is required by 10 CFR 50, Appendix A, GDC 54, "Piping Systems Penetrating Containment" (Ref. 1), to ensure a water seal to a specific class of containment isolation valves (double disc gate valves) during a LOCA, to prevent leakage of containment atmosphere through the gate valves.

The CVIWS is designed to inject water between the two seating surfaces of double disc gate valves used for Containment isolation. The injection pressure is higher than Containment design peak pressure during a LOCA. This will prevent leakage of the Containment atmosphere through the gate valves, thereby reducing potential offsite dose below the values specified by 10 CFR 50.67 limits following the postulated accident.

During normal power operation, the system is in a standby mode and does not perform any function. During accident situations the CVIWS is activated to perform its safety related function, thus limiting the release of containment atmosphere past specific containment isolation valves, in order to mitigate the consequences of a LOCA. Containment isolation valves, for systems which are not used to mitigate the consequences of an accident, will be supplied with CVIWS seal water upon receipt of a Phase A isolation signal. Containment isolation valves, for accident mitigating systems which are supplied with seal water from the CVIWS, have their seal water supplies actuated by a Containment Pressure -

High-High signal.

The system consists of two independent, redundant trains; one supplying gate valves that are powered by the A train diesel and the other supplying gate valves powered by the B train diesel. This separation of trains prevents the possibility of both containment isolation valves not sealing due to a single failure.

Each train consists of a surge chamber which is filled with water and pressurized with nitrogen. One main header exits the chamber and splits into several headers. A solenoid valve is located in the main header before any of the branch headers which will open after a 60 second delay on a Phase A isolation signal. Each of the headers supply injection water Catawba Units 1 and 2 B 3.6.17-1 Revision No. 1

CVIWS B 3.6.17 BASES BACKGROUND (continued) to containment isolation valves located in the same general location, and close on the same engineered safety signal. A solenoid valve is located in each header which supplies seal water to valves closing on a Containment Pressure - High-High signal. These solenoid valves open after a 60 second delay on a Containment Pressure - High-High signal.

Since a Phase A isolation signal occurs before a Containment Pressure -

High-High signal, the solenoid valve located in the main header will already be injecting water to'Containment isolation valves closing on a Phase A isolation signal. This leaves an open path to the headers supplying injection water on a Containment Pressure - High-High signal.

The delay for the solenoid valves opening is to allow adequate time for the slowest gate valve to close, before water is injected into the valve seat.

Makeup water is provided from the Demineralized Water Storage Tank for testing and adding water to the surge chamber during normal plant operation. Assured-water is provided from the essential header of the Nuclear Service Water System (NSWS). This supply is assured for at least 30 days following a postulated accident. If the water level in the surge chamber drops below the low-low level or if the surge chamber nitrogen pressure drops below the low-low pressure after a Phase A isolation signal, asolenoid valve in the supply line from the NSWS will automatically open and remains open, assuring makeup to the CVIWS at a pressure greater than 110% of peak Containment accident pressure.

Overpressure protection is provided to relieve the pressure buildup caused by the heatupof a trapped volume of incompressible fluid between two positively closing valves (due to containment temperature transient) back into containment where an open relief path exists.

APPLICABLE The CVIWS design basis is established by the consequences of the SAFETY ANALYSES limiting DBA, which is a LOCA. The accident analysis (Ref. 2) assumes that only one train of the CVIWS is functional due to a single failure that disables the other train. Makeup water can be assured from the NSWS for 30 days following a postulated LOCA.

The CVIWS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 3).

Catawba Units 1 and 2 B 3.6.17-2 IRevision No. 0

CVIWS B 3.6.17 BASES REFERENCES 1. 10 CFR 50, Appendix A, GDC 54.

2. UFSAR, Section 6.2.

3. ' 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. 10 CFR 50.67.

Catawba Units 1 and 2 B 3.6.17-5 Revision No. 1

MSIVs B 3.7.2 BASES LCO This LCO requires that four MSIVs in the steam lines be OPERABLE.

The MSIVs are considered OPERABLE when the isolation times are within limits, and they close on an isolation actuation signal.

This LCO provides assurance that the MSIVs will perform their design safety function to mitigate the consequences of accidents that could result in offsite exposures comparable to the 10 CFR 50.67 (Ref. 5) limits or the NRC staff approved licensing basis.

APPLICABILITY The MSIVs must be OPERABLE in MODE 1, and in MODES 2 and 3 except when closed and de-activated, when there is significant mass and energy in the RCS and steam generators. When the MSIVs are closed, they are already performing the safety function.

In MODE 4, normally most of the MSIVs are closed, and the steam generator energy is low.

In MODE 5 or 6, the steam generators do not.contain much energy because their temperature is below the boiling point of water; therefore, the MSIVs are not required for isolation of potential high energy secondary system pipe. breaks in these MODES.

ACTIONS A.1 With one MSIV inoperable in MODE 1, action must be taken to restore OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Some repairs to the MSIV can be made with the unit hot. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable, considering the low probability of an accident occurring during this time period that would require a closure of the MSIVs.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is greater than that normally allowed for containment isolation valves because the MSIVs are valves that isolate a closed system penetrating containment. These valves differ from other containment isolation valves in that the closed system provides an additional means for containment isolation.

B.1 If the MSIV cannot be restored to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Condition C would be entered. The Completion Times are Catawba Units 1 and 2 B 3.7.2-3 Revision No. 1

MSIVs B 3.7.2 BASES ACTIONS (continued) reasonable, based on operating experience, to reach MODE 2 and to close the MSIVs in an orderly manner and without challenging unit systems.

C.1 and C.2 Condition C is modified by a Note indicating that separate Condition entry is allowed for each MSIV.

Since the MSIVs are required to be OPERABLE in MODES 2 and 3, the inoperable MSIVs may either be restored to OPERABLE status or closed.

When closed, the MSIVs are already in the position required by the assumptions in the safety analysis.

The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is consistent with that allowed in Condition A.

For inoperable MSIVs that cannot be restored to OPERABLE status within the specified Completion Time, but are closed, the inoperable MSIVs must be verified on a periodic basis to be closed. This is necessary to ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, in view of MSIV status indications available in the control room, and other administrative controls, to ensure that these Valves are in the closed position.

D.1 and D.2 If the MSIVs cannot be restored to OPERABLE status or are not closed within the associated Completion Time, the unit must be placed in a MODE. in which the LCO does not apply. To achieve this status, the unit must be placed at least in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from MODE 2 conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that MSIV closure time is < 8.0 seconds on an actual or simulated actuation signal. The MSIV closure time is assumed in the accident and containment analyses. This Surveillance is normally Catawba Units 1 and 2 B 3.7.2-4 R&vision No. 0

MSIVs B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued) performed upon returning the unit to operation following a refueling outage. The MSIVs should not be tested at power, since even a part stroke exercise increases the risk of a valve closure when the unit is generating power. As the MSIVs are not tested at power, they are exempt fromthe ASME Code, Section Xl (Ref. 6), requirements during.

operation in MODE 1 or 2. The Frequency is in accordance with the Inservice Testing Program.

This test is conducted in MODE 3 with the unit at operating temperature and pressure, as discussed in Reference 6 exercising requirements. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

REFERENCES 1. UFSAR, Section 10.3.

2. UFSAR, Section 6.2.

3. UFSAR, Section 15.1.5.

4. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

5. 10 CFR 50.67.

6. ASME, Boiler and Pressure Vessel Code, Section Xl.

Catawba Units 1 and 2 B 3.7.2-5 Revision No. 1

CSS B 3.7.6 BASES ACTIONS (continued)

MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance on the steam generator for heat removal, within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that the CSS contains the required inventory of cooling water. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on operating experience and the need for operator awareness of unit evolutions that may affect the CSS inventory between checks. Also, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate in view of other indications in the control room, including alarms, to alert the operator to abnormal deviations in the CSS level.

REFERENCES 1. UFSAR, Section 10.4.

2. UFSAR, Chapter 6.

3. UFSAR, Chapter 15.

Catawba Units 1 and 2 B 3.7.6-3 Revision No. 2

CRAVS B 3.7.10 B 3.7 PLANT SYSTEMS B 3.7.10 Control Room Area Ventilation System (CRAVS)

BASES BACKGROUND The CRAVS ensures that the control room will remain habitable for personnel during and following all credible accident conditions. This function is accomplished-by pressurizing the control room to > 1/8 (0.125) inch water gauge with respect to all surrounding areas, filtering the outside air used for pressurization, and filtering a portion of the return air from the control room to clean up the control room environment.

The CRAVS consists of two independent, redundant trains of equipment.

Each train consists of:

a pressurizing filter train fan (1 CRA-PFTF-1 or 2CRA-PFTF-1)

a filter unit (1 CRA-PFT-1 or 2CRA-PFT-1) which includes moisture separator/prefilters, HEPA filters, and carbon adsorbers

the associated ductwork, dampers/valves, and controls Inherent in the CRAVS ability to pressurize the control room is the control room pressure boundary. This pressure boundary includes: (1) the control room walls, floor, roof, doors, and all penetrations of those, (2) any piping or ductwork which penetrates into the control room, and (3) the control room ventilation system proper consisting of ductwork, filter units, dampers, and fans. These boundaries must be intact or properly isolated for the CRAVS to function properly.

The CRAVS can be operated either manually or automatically. Key operated selector switches located in the control room initiate operation of all train related CRAVS equipment. The selected train is in continuous operation. Outside air for pressurization and makeup to the control room is supplied from two independent intakes. This outside air is mixed with return air from the control room before being passed through the filter unit. In the filter unit, moisture separator/prefilters remove any large particles in the air, and any entrained water droplets present. A HEPA filter bank upstream of the carbon adsorber filter bank functions to remove particulates and a second bank of HEPA filters follow the carbon adsorber to collect carbon fines. Only the upstream HEPA filters and carbon adsorber bank are credited in the analysis. A heater is included within each filter train to reduce the relative humidity of the airstream, although no credit is taken in the safety analysis. The heaters are not required for OPERABILITY since the carbon laboratory tests are performed at 95% relative humidity, but have been maintained in the Catawba Units 1 and 2 B 3.7.10-1 Revision No. 2

CRAVS B 3.7.10 BASES BACKGROUND (continued) system to provide additional margin (Ref. 9). Continuous operation of each train for at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per month, with the heaters on, reduces moisture buildup on the HEPA filters and adsorbers.

Upon receipt of an Engineered Safety Feature (ESF) signal, the selected CRAVS train.continues to operate and-the pressurizing filter train fan of the non-selected train is started. This assures control room pressurization, assuming an active failure of one of the pressurizing filter train fans.

The outside air for pressurization is continuously monitored for the presence of smoke, radiation, or chlorine by, non-safety related detectors.

If smoke, radiation, or chlorine is detected in an outside air intake, an alarm is received in the control.room, alerting the operators of this condition. The operator will take the required action to close the affected intake, if necessary, per the guidance of the Annunciator Response Procedures.

A single CRAVS train is capable of pressurizing the control room to greater than or equal to 0.125 inches water gauge. The CRAVS is designed in accordance with Seismic Category 1 requirements. The CRAVS operation in maintaining the control room habitable is discussed in the UFSAR, Sections 6.4 and 9.4.1 (Refs. 1 and 2).

The CRAVS is designed to maintain the control room environment for 30 days of continuous occupancy after a Design Basis Accident (DBA) without the total effective dose equivalent in the control room exceeding 5 rem.

APPLICABLE The CRAVS components are arranged in redundant, safety related SAFETY ANALYSES ventilation trains. The CRAVS provides airborne radiological protection -

for the control room operators, as demonstrated by the control room accident dose analyses for the most limiting design basis loss of coolant accident, fission product release presented in the UFSAR, Chapter 15 (Ref. 3).

The analysis of toxic gas releases demonstrates that the toxicity limits are not exceeded in the control room following a toxic chemical release, as presented in Reference 1.

The worst case single active failure of a component of the CRAVS,

assuming a loss of offsite power, does not impair the ability of the system to perform its design function.

Catawba Units 1 and 2 B 3.7.10-2 Revision No. 4

CRAVS B 3.7.10 BASES APPLICABLE SAFETY ANALYSES (continued)

The CRAVS satisfies Criterion 3 of 10 CFR 50.36 (Ref. 4).

LCO Two independent and redundant CRAVS trains are required to be OPERABLE to ensure that at least one is available assuming a single failure disables the other train. Total system failure could result in the total effective dose equivalent to the control room operator exceeding 5 rem in the event of a large radioactive release.

The CRAVS is considered OPERABLE when the individual components necessary to limit operator exposure are OPERABLE in both trains. A CRAVS train is OPERABLE when the associated:

a. Pressurizing filter train fan is OPERABLE;

b. HEPA filters and carbon adsorbers are not excessively restricting flow, and are capable of performing their filtration functions; and

c. Ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

In addition, the control room pressure boundary must be maintained, including the integrity of the walls, floors, roof, ductwork, and access doors.

The CRAVS is shared between the two units. The system must be OPERABLE for each unit when that unit is in the MODE of Applicability.

Additionally, both normal and emergency power must also be OPERABLE because the system is shared. If a CRAVS component becomes inoperable, or normal or emergency power to a CRAVS component becomes inoperable, then the Required Actions of this LCO must be entered independently'for each unit that is in the MODE of applicability of the LCO.

The LCO is modified by a Note allowing the control room pressure boundary to be opened intermittently under administrative controls.. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for control room pressure boundary isolation is indicated.

Catawba Units 1 and 2 8 3.7.10-3 Revision No. 6

CRAVS B 3.7.10 BASES APPLICABILITY In MODES 1, 2, 3, 4, 5, and 6, CRAVS must be OPERABLE to control operator exposure during and following a DBA.

During movement of irradiated fuel assemblies, the CRAVS must be OPERABLE to cope with the release from a fuel handling accident.

ACTIONS A.1 When one CRAVS train is inoperable in MODES 1,2,3,4,5,or 6, action must be taken to restore OPERABLE status within 7 days. In this Condition, the remaining OPERABLE CRAVS train is adequate to perform the control room protection function. However, the overall reliability is reduced because a single failure in the OPERABLE CRAVS train could result in loss of CRAVS function.. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.

8.1 If the control room pressure boundary is inoperable in MODES 1, 2, 3, or 4 such that the CRAVS trains cannot establish or maintain the required pressure, action must be taken to restore an OPERABLE control room pressure boundary within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . During the period that the control room pressure boundary is inoperable, appropriate compensatory measures (consistent with the intent of GDC 19) should be utilized to protect control room operators from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable based on the low probability of a DBA occurring during this time period and the use of compensatory measures. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the control room pressure boundary.

C.1 and C.2 In MODE 1, 2, 3, or 4, if the inoperable CRAVS or control room pressure boundary train cannot be restored to OPERABLE status within the required Completion Time, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Catawba Units 1 and 2 B 3.7.10-4 Revision No. 4

Spent Fuel Pool Water Level B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the spent fuel pool design is given in the UFSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling System is, given in the UFSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the UFSAR, Section 15.7.4 (Ref. 3).

APPLICABLE The minimum water level in the spent fuel pool meets the assumptions SAFETY ANALYSES of the fuel handling accident described in Regulatory Guide 1.183 Appendix B (Ref. 4).. The resultant total effective dose equivalent is within the acceptance criteria of 10 CFR 50.67 (Ref. 5).

According to Reference 4, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage. racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The spent fuel pool water level satisfies Criterion 2 of 10 CFR 50.36 (Ref.

6).

Catawba Units 1 and 2 B 3.7.14-1 Revision No. 1

Spent Fuel Pool Water Level B 3.7.14 BASES LCO The spent fuel pool water level is required to be > 23 ft over the top of irradiated fuel assemblies seated in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 3). As such, it is the minimum required for fuel storage and movement within the spent fuel pool.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in.the spent fuel pool, since the potential for a release of fission products exists.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel pool water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pool is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies sufficient spent fuel pool water is available in the event of a fuel handling accident. The water level in the spent fuel pool must be checked periodically. The 7 day Frequency is appropriate because the volume in the pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the spent fuel pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.6.1.

Catawba Units 1 and 2 B 3.7.14-2 -Revision No. 0

Spent Fuel Pool Water Level B 3.7.14 BASES REFERENCES 1. UFSAR, Section 9.1.2.

2. UFSAR, Section 9.1.3.

3. UFSAR, Section 15.7.4.

4. Regulatory Guide 1.183, Appendix B.

5. 10 CFR 50.67.

6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

Catawba Units 1 and 2 B 3.7.14-3 Revision No. 1

Secondary Specific Activity B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Reactor Coolant System (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives and, thus, indicates current conditions. During transients, 1-131 spikes have been observed as well as increased releases of some noble gases.

Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and 'the reactor coolant LEAKAGE. Most of the iodine isotopes have short half lives, (i.e., < 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ). 1-13.1, with a half life of 8.04 days, concentrates faster than it decays, but does not reach equilibrium because of blowdown and other losses.

Operating a unit at the allowable limits will result in a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months EAB exposure of less than the acceptance criteria of 10 CFR 50.67 (Ref. 1) and Regulatory Guide 1.183 (Ref. 4) limits.

APPLICABLE The accident analysis of the main steam line break (MSLB), as SAFETY ANALYSES discussed in the UFSAR, Chapter 15 (Ref. 2) assumes the initial secondary coolant specific activity to have a radioactive isotope concentration of 0.10 pCi/gm DOSE EQUIVALENT 1-131. This assumption is used in the analysis for determining the radiological consequences of-the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed the acceptance criteria of the unit EAB limits (Ref. 1, 4) for total effective dose equivalent rates.

With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the MSSVs and steam generator power operated Catawba Units 1 and 2 B 3.7.17-1 Revision No. 1

Secondary Specific Activity B 3.7.17.

BASES APPLICABLE SAFETY ANALYSES (continued) relief valves (SG PORVs). The Auxiliary Feedwater System supplies the necessary makeup to the steam generators. Venting continues until the reactor coolant temperature and pressure have decreased sufficiently for the Residual Heat Removal System to complete the cooldown.

In the evaluation of the radiological consequences of this accident, the activity released from the steam generator connected to the failed steam line is assumed to be released directly to the environment. The unaffected steam generator isassumed to discharge steam and any entrained activity through the MSSVs and SG PORVs during the event.

Since no credit is taken in the analysis for activity plate out or retention, the resultant radiological consequences represent a conservative estimate of the potential integrated dose due to the postulated steam line failure.

Secondary specific activity limits satisfy Criterion 2 of 10 CFR 50.36 (Ref.

3).-

LCO As indicated .in the Applicable Safety Analyses, the specific activity of the secondary coolant is required to be

0.10 pCi/gm DOSE EQUIVALENT 1-131 to limit the radiological consequences of a Design Basis Accident (DBA) to within the acceptance criteria (Ref. 1, 4).

Monitoring the specific activity of the secondary coolant-ensures that when secondary specific activity limits are exceeded, appropriate actions are taken in a timely manner to place the unit in an operational MODE that would minimize the radiological consequences of a DBA.

APPLICABILITY In MODES 1., 2, 3, and 4, the limits on secondary specific activity apply due to the potential for secondary steam releases to the atmosphere.

In MODES 5 and 6, the steam generators are not being used for heat removal. Both the RCS and steam generators are depressurized, and primary to secondary LEAKAGE is minimal. Therefore, monitoring of secondary specific activity is not required.

Catawba Units 1 and 2 B 3.7.17-2 Revision No. 1

Secondary Specific Activity B 3.7.17 BASES ACTIONS A.1 and A.2 DOSE EQUIVALENT 1-131 exceeding the allowable value in the secondary coolant, is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity cannot be restored to within limits within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. The 31 day Frequency is based on the detection of increasing trends of the level of DOSE EQUIVALENT 1-131, and allows for appropriate action to be taken to maintain levels below the LCO limit.

REFERENCES 1. 10 CFR 50.67.

2. UFSAR, Section 15.1.5.

3. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

4. Regulatory Guide 1.183, July 2000.

Catawba Units 1 and 2 B 3.7.17-3 Revision No. 1

Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGROUND During movement of recently irradiated fuel assemblies (i.e., fuel assemblies that have occupied part of a critical reactor core within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LCO requirements are met. In MODES 1, 2, 3,

and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment." In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be. less stringent. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained within the acceptance criteria of 10 CFR 50.67 and Regulatory Guide 1.183. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. During movement of recently irradiated fuel assemblies within containment, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for Catawba Units 1 and 2 B 3.9.3-1 Revision No. 3

Containment Penetrations B 3.9.3 BASES BACKGROUND (continued) extended periods when frequent containment entry is necessary. During movement of recently irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed.

The requirements for containment penetration closure ensure that a release of fission product radioactivity within containment will be restricted from escaping to the environment. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident involving recently irradiated fuel during refueling.

The Containment Purge Exhaust System includes two trains. Purge air is exhausted from the containment through the Containment Purge Exhaust System to the unit vent where it is monitored for radioactivity level by the unit vent monitor prior to release to the atmosphere. The Containment Purge Exhaust System consists of two 50 percent capacity filter trains and fans. There is one purge exhaust duct penetration through the Reactor Building wall from the annulus area. There are three purge exhaust penetrations through the containment vessel, two from the upper compartment and one from the lower compartment. Two normally closed isolation valves at each penetration through the containment vessel provide containment isolation. One normally closed isolation damper at the Reactor Building wall provides annulus isolation.

The upper compartment purge exhaust ductwork is arranged to draw exhaust air into a plenum around the periphery of the refueling canal, effecting a ventilation sweep of the canal during the refueling process.

The lower compartment purge exhaust ductwork is arranged so as to sweep the reactor well during the refueling process.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment, penetrations during recently irradiated fuel movements.

APPLICABLE During movement of recently irradiated fuel assemblies within I SAFETY ANALYSES containment, the most severe radiological consequences result from a I fuel handling accident involving recently irradiated fuel. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 1). Fuel handling accidents, analyzed in Reference 2, include Catawba Units 1 and 2 B 3.9.3-2 Revision No. 1 1

Containment Penetrations B 3.9.3 BASES APPLICABLE SAFETY ANALYSES (continued) dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The requirements of LCO 3.9.6, "Refueling Cavity Water Level," and the minimum decay time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months without containment closure capability ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are within the guideline values specified in 10 CFR 50.67 and Regulatory Guide 1.183.

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).

LCO This LCO limits the consequences of a fuel handling accident involving handling recently irradiated fuel in containment by limiting the potential escape paths for fission product radioactivity released within containment.

The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for penetrations exhausting through an OPERABLE Containment Purge Exhaust System HEPA filter and carbon adsorber during movement of recently irradiated fuel assemblies.

APPLICABILITY The containment penetration requirements are applicable during movement of recently irradiated fuel assemblies within containment because this is when there is a potential for the limiting fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when movement of recently irradiated fuel assemblies within containment is not being conducted, the potential for a limiting fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.

During movement of recently irradiated fuel assemblies, ventilation system and radiation monitor availability (as defined by NUMARC 91-06) should -

be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the RCS decays fairly rapidly.

The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay, and to avoid unmonitored releases.

A single normal or contingency method to promptly close primary or secondary containment penetrations exists. Such prompt methods need Catawba Units 1 and.2 B 3.9.3-3 Revision No. 3

Containment Penetrations B 3.9.3 BASES APPLICABILITY (continued) not completely block the penetration or be capable of resisting pressure.

The purpose is to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper directions such that it can be treated and monitored.

ACTIONS A.1 and A.2 If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere.

to the outside atmosphere is not in the required status, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of recently irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

B.1 and B.2 With one or more Containment Purge Exhaust System heaters inoperable, the heater must be restored to OPERABLE status within 7 days. Alternatively, a report-must be initiated per Specification 5.6.6, which details the reason for the heater's inoperability and the corrective action required to return the heater to OPERABLE status.

The heaters do not affect OPERABILITY of the Containment Purge Exhaust System filter trains because carbon adsorber efficiency testing is performed at 30°C and 95% -relative humidity. The accident analysis shows that site boundary radiation doses are within the limits of 10 CFR 50.67 and Regulatory Guide 1.183 during a DBA LOCA under these conditions.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to.be in its closed position is -in that position. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are exhausting through an OPERABLE Containment Purge Exhaust System HEPA Filter and carbon adsorber.

The Surveillance is performed every 7 days during movement of recently irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. As such, this Surveillance ensures that a postulated fuel handling accident involving recently irradiated fuel that Catawba Units 1 and 2 B 3.9.3-4 Revision No. 3

Refueling Cavity Water Level B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies or performance of CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to < 25% of 10 CFR 50.67 limits, as provided by the guidance of Reference 3.

APPLICABLE During CORE ALTERATIONS and movement of.irradiated fuel SAFETY ANALYSES assemblies, the water level in the refueling canal and the refueling cavity is an initial condition .design parameter in the analysis of a fuel handling accident in containment, as postulated by Regulatory Guide 1.183 (Ref. 1). A minimum water level of 23 ft (Appendix B, Section 2) allows a decontamination factor of 200 (Appendix B, Section 2) to be used in the accident analysis for iodine. This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 5% of the total fuel rod iodine inventory with the exception of 8% Iodine-1 31 (Ref. 1).

The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 4 and 5).

Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36 (Ref. 6).

LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits, as provided by the guidance of Reference 3.

Catawba Units 1 and 2 B 3.9.6-1 Revision No. 1

Refueling Cavity Water Level B 3.9.6 BASES APPLICABILITY LCO 3.9.6 is applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, and is also applicable when moving irradiated fuel assemblies within containment. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident.

Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.14, "Fuel Storage Pool Water Level."

ACTIONS A.1 and A.2 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.

The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.

Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.

Catawba Units 1 and 2 B 3.96-2 - Revision No. 0

Refueling Cavity Water Level B 3.9.6 BASES REFERENCES 1. Regulatory Guide 1.183, July 2000.

2. UFSAR, Section 15.7.4.

3. NUREG-0800, Section 15.7.4.

4. 10 CFR 50.67.

5. Malinowski, D. D., Bell, M. J., Duhn, E., and Locante, J., WCAP-828, Radiological Consequences of a Fuel Handling Accident, December 1971.

6. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

Catawba Units 1 and 2 B 3.9.6-3 Revision No. 1

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