Text

Technical Specification Bases Changes
	ML17181A170
Person / Time
Site:	Catawba
Issue date:	06/27/2017
From:	Simril T; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CNS-17-033
	Download: ML17181A170 (98)
	v • d • e

Tom Simril e(.,DUKE Vice President

~ ENERG'l Catawba Nuclear Station Duke Energy CN01VP 14800 Concord Road York, SC 29745 o: 803. 701.4251 f: 803. 701.3221 tom.simril@duke-energy.com CNS-17-033 10CFR 50.4 June 27, 2017 U.S. Nuclear Regulatory Commission Document Control Desk

Washington, DC 20555-0001

Subject:

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 Technical Specification 5.5.14, "Technical Specifications (TS) Bases Control Program."

Any questions regarding this information should be directed to Tolani Owusu, Regulatory Affairs, at (803) 701-5385.

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

Tom Simril Vice President, Catawba Nuclear Station Attachments: 1) Insertion/Removal Instructions

2) Replacement Pages WWW.duke-energy.com

Catawba Nuclear Station Technical Specifications Manual Amendments CNS-17-033 June 27, 2017 Page 2 xc: w/attachments Catherine Haney, Regional Administrator U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 Mike Mahoney NRC Project Manager (CNS)

U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 0 - 8H4A 11555 Rockville Pike Rockville, MD 20852-27 46 Joseph Austin, Senior Resident Inspector Catawba Nuclear Station

Removal and insertion instructions for Catawba Nuclear Station Tech Spec Bases Changes for January 1 - June 1, 2017.

REMOVE THESE PAGES INSERT THESE PAGES LIST OF EFFECTIVE PAGES Pages 1-19 Pages 1-19 Revision 10 Revision 11 TECHNICAL SPECIFICATIONS BASES TAB 3.4 B 3.4.6 B 3.4.6-5 B 3.4.6 B 3.4.6-6 Revision 4 Revision 5 B 3.4.7 B 3.4.7-5 B 3.4.7-1-B3.4.7-7 Revision 6 Revision 7 B 3.4.8 B 3.4.8-3 B 3.4.8-1-B3.4.8-4 Revision 3 Revision 4 BASES TAB 3.5 B 3.5.1-1-B 3.5.1-8 B 3.5.1-1-B3.5.1-8 Revision 3 Revision 4 B 3.5.2 B 3.5.2-10 B 3.5.2 B 3.5.2-11 Revision 3 Revision 4 B 3.5.3 B 3.5.3-3 B 3.5.3-1 "'- B 3.5.3-3 Revision 1 Revision 2 BASES TAB 3.6 B 3.6.3-1-B3.6.3-14 B-3.6.3-1-B3.6.3-14 Revision 4 Revision 5

B 3.6.6-1-B3.6.6-7 B 3.6.6-1-B3.6.6-8 Revision 6 Revision 7 BASES TAB 3.9 B 3.9.2 B 3.9.2-3 B 3.9.2 B 3.9.2-3 Revision 5 Revision 6 B 3.9.4 B 3.9.4-4 B 3.9.4 B 3.9.4-6 Revision 4 Revision 5 B 3.9.5 B 3.9.5-4 B 3.9.5 B 3.9.5-5 Revision 3 Revision '4 If you have any questions concerning the contents of this Technical Specification Bases update, please contact Toni Lowery at (803) 701-5046.

Catawba Nuclear Station Technical Specifications Page Number List of Effective Pages Amendments Revision Date 177/169 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 1.1-2 268/264 6/25/12 1.1-3 268/264 6/25/12 1.1-4 268/264 6/25/12 1.1-5 281/277 4/29/16 1.1-6 268/264 6/25/12 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 Catawba Units 1 and 2 Page 1 5/4/17

1.4-4 2.0-1 3.0-1 3.0-2 173/165 210/204 235/231 235/231 9/30/98 12/19/03 3/19/07 3/19/07 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 3.1.1-1 263/259 3/29/11 3.1.2-1 173/165 9/30/98 3.1.2-2 263/259 3/29/11 3.1.3-1 173/165 9/30/98 3.1.3-2 275/271 04/14/15 3.1.3-3 173/165 9/30/98 3.1.4-1 173/165 9/30/98 3.1.4-2 173/165 . 9/30/98 3.1.4-3 263/259 3/29/11 3.1.4-4 263/259 3/29/11 3.1.5-1 173/165 9/30/98 3.1.5-2 263/259 3/29/11 3.1.6-1 173/165 9/30/98 3.1.6-2 173/165 9/30/98 3.1.6-3 263/259 3/29/11 3.1.7-1 173/165 9/30/98 3.1.7-2 173/165 9/30/98 3.1.8-1 173/165 9/30/98 3.1.8-2 263/259 3/29/11 3.2.1-1 173/165 9/30/98 3.2.1-2 173/165 9/30/98 3.2.1-3 263/259 3/29/11 3.2.1-4 263/259 3/29/11 3.2.1-5 263/259 3/29/11 3.2.2-1 173/165 9/30/98 3.2.2-2 173/165 9/30/98 Catawba Units 1 and 2 Page 2 5/4/17

3.2.2-3 3.2.2-4 3.2.3-1 3.2.4-1 263/259 263/259 263/259 173/165 3/29/11 3/29/11 3/29/11 9/30/98 3.2.4-2 173/165 9/30/98 3.2.4-3 173/165 9/30/98 3.2.4-4 263/259 3/29/11 3.3.1-1 173/165 9/30/98 3.3.1-2 247/240 12/30/08 3.3.1-3 247/240 12/30/08 3.3.1-4 207/201 7/29/03 3.3.1-5 247/240 12/30/08 3.3.1-6 247/240 12/30/08 3.3.1-7 247/240 12/30/08 3.3.1-8 173/165 9/30/98 3.3.1-9 263/259 3/29/11 3.3.1-10 263/259 3/29/11 3.3.1-11 263/259 3/29/11 3.3.1-12 278/274 4/08/16 3.3.1-13 263/259 3/29/11 3.3.1-14 263/259 3/29/11 3.3.1-15 263/259 3/29/11 3.3.1-16 278/274 4/08/16 3.3.1-17 263/259 3/29/11 3.3.1-18 263/259 3/29/11 3.3.1-19 278/274 4/08/16 3.3.1-20 263/259 3/29/11 3.3.1-21 263/259 3/29/11 3.3.1-22 263/259 3/29/11 3.3.2-1 173/165 9/30/98 3.3.2-2 247/240 12/30/08 3.3.2-3 247/240 12/30/08 3.3.2-4 247/240 12/30/08 3.3.2-5 264/260 6/13/11 Catawba Units 1 and 2 Page 3 5/4/17

3.3.2-6 3.3.2-7 3.3.2-8 3.3.2-9 264/260 249/243 249/243 249/243 6/13/11 4/2/09 412109 4/2/09 3.3.2-10 263/259 3/29/11 3.3.2-11 263/259 3/29/11 3.3.2-12 263/259 3/29/11 3.3.2-13 277/273 12/18/15 3.3.2-14 277/273 12/18/15 3.3.2-15 277/273 12/18/15 3.3.2-16 277/273 12/18/15 3.3.2-17 277/273 12/18/15 3.3.2-18 (new) 277/273 12/18/15 3.3.3-1 219/214 3/1/05 3.3.3-2 219/214 3/1/05 3.3.3-3 263/259 3/29/11 3.3.3-4 219/214 3/1/05 3.3.4-1 213/207 4/29/04 3.3.4-2 263/259 3/29/11 3.3.4-3 272/268 2/27/14 3.3.5-1 173/165 9/30/98 3.3.5-2 277/273 12/18/15 3.3.6-1 196/189 3120102 3.3.6-2 263/259 3/29/11 3.3.6-3 196/189 3120102 3.3.9-1 207/201 7/29/03 3.3.9-2 207/201 7/29/03 3.3.9-3 263/259 3/29/11 3.3.9-4 263/259 3/29/11 3.4.1-1 210/204 12/19/03 3.4.1-2 210/204 12/19/03 3.4.1-3 263/259 3/29/11 3.4.1-4 283/279 6/02/16 3.4.1-5 (deleted) 184/176 3/01/00 Catawba Units 1 and 2 Page4 5/4/17

3.4.1-6 (deleted) 3.4.2-1 3.4.3-1 3.4.3-2 184/176 173/165 173/165 263/259 3/01/00 9/30/98 9/30/98 3/29/11 3.4.3-3 281/277 4/29/16 3.4.3-4 212/206 314104 3.4.3-5 281/277 4/29/16 3.4.3-6 212/206 3/4/04 3.4.4-1 263/259 3/29/11 3.4.5-1 207/201 7/29/03 3.4.5-2 207/201 7/29/03 3.4.5-3 263/259 3/29/11 3.4.6-1 212/206 314104 3.4.6-2 263/259 3/29/11 3.4.6-3 282/278 4/26/17 3.4.7-1 212/206 314104 3.4.7-2 263/259 3/29/11 3.4.7-3 282/278 4/26/17 3.4.8-1 207/201 7/29/03 3.4.8-2 282/278 4/26117 3.4.9-1 173/165 9/30/98 3.4.9-2 263/259 3/29/11 3.4.10-1 212/206 314104 3.4.10-2 173/165 9/30/98 3.4-11-1 213/207 4/29/04 3.4.11-2 173/165 9/30/98 3.4.11-3 263/259 3/29/11 3.4.11-4 263/259 3/29/11 3.4.12-1 212/206 314104 3.4.12-2 213/207 4/29/04 3.4.12-3 212/206 314104 3.4.12-4 212/206 314104 3.4.12-5 263/259 3/29/11 3.4.12-6 263/259 3/29/11 Catawba Units 1 and 2 Page 5 5/4/17

3.4.12-7 3.4.12-8 3.4.13-1 3.4.13-2 263/259 263/259 267/263 267/263 3/29/11 3/29/11 3/12/12 3/12/12 3.4.14-1 173/165 9/30/98 3.4.14-2 173/165 9/30/98 3.4.14-3 263/259 3/29/11 3.4.14-4 263/259 3/29/11 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 3.4.15-4 263/259 3/29/11 3.4.16-1 268/264 6/25/12 3.4.16-2 268/264 6/25/12 3.4.16-3(deleted) 268/264 6/25/12 3.4.16-4(deleted) 268/264 6/25/12 3.4.17-1 263/259 3/29/11 3.4.18-1 280/276 4/26/16 3.4.18-2 280/276 4/26/16 3.5.1-1 211/205 12/23/03 3.5.1-2 263/259 3/29/11 3.5.1-3 263/259 3/29/11 3.5.2-1 253/248 10/30/09 3.5.2-2 282/278 4/26/17 3.5.2-3 263/259 3/29/11 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 3.5.4-2 269/265 7/25/12 3.5.5-1 173/165 9/30/98 3.5.5-2 263/259 3/29/11 3.6.1-1 173/165 9/30/98 3.6.1-2 192/184 7/31/01 3.6.2-1 173/165 9/30/98 Catawba Units 1 and 2 Page6 5/4/17

3.6.2-2 3.6.2-3 3.6.2-4 3.6.2-5 173/165 173/165 173/165 263/259 9/30/98 9/30/98 9/30/98 3/29/11 3.6.3-1 173/165 9/30/98 3.6.3-2 173/165 9/30/98 3.6.3-3 173/165 9/30/98 3.6.3-4 173/165 9/30/98 3.6.3-5 263/259 3/29/11 3.6.3-6 263/259 3/29/11 3.6.3-7 192/184 7/31/01 3.6.4-1 263/259 3/29/11 3.6.5-1 173/165 9/30/98 3.6.5-2 263/259 3/29/11 3.6.6-1 282/278 4/26/17 3.6.6-2 282/278 4/26/17 3.6.8-1 213/207 4129104

3.6.8-2 3.6.9-1 3.6.9-2 3.6.10-1 263/259 253/248 263/259 173/165 3/29/11 10/30/09 3/29/11 9/30/98 3.6.10-2 263/259 3/29/11

. 3.6.11-1 263/259 3/29/11 3.6.11-2 263/259 3/29/11 3.6.12-1 263/259 3/29/11 3.6.12-2 263/259 3/29/11 3.6.12-3 263/259 3/29/11 3.6.13-1 256/251 6/28/10 3.6.13-2 263/259 3/29/11 3.6.13-3 263/259 3/29/11 3.6.14-1 173/165 9/30/98 3.6.14-2 263/259 3/29/11 3.6.14-3 270/266 8/6/13 3.6.15-1 173/165 9/30/98 Catawba Units 1 and 2 Page? 5/4/17

3.6.15-2 3.6.16-1 3.6.16-2 3.6.17-1 263/259 263/259 263/259 253/248 3/29/11 3/29/11 3/29/11 10/30/09 3.7.1-1 173/165 9/30/98 3.7.1-2 173/165 9/30/98 3.7.1-3 281/277 4/29/16 3.7.2-1 173/165 9/30/98 3.7.2-2 244/238 9/08/08 3.7.3-1 173/165 9/30/98 3.7.3-2 244/238 9/08/08 3.7.4-1 213/207 4/29/04 3.7.4-2 263/259 3/29/11 3.7.5-1 253/248 10/30/09 3.7.5-2 173/165 9/30/98 3.7.5-3 263/259 3/29/11 3.7.5-4 263/259 3/29/11 3.7.6-1 173/165 9/30/98 3.7.6-2 263/259 3/29/11 3.7.7-1 253/248 10/30/09 3.7.7-2 263/259 3/29/11 3.7.8-1 271/267 08/09/13 3.7.8-2 271/267 08/09/13 3.7.8-3 271/267 08/09/13 3.7.8-4 271/267 08/09/13 3.7.9-1 263/259 3/29/11 3.7.9-2 263/259 3/29/11 3.7.10-1 250/245 7130109 3.7.10-2 260/255 8/9/10 3.7.10-3 263/259 3/29/11 3.7.11-.1 198/191 4/23/02 3.7.11-2 263/259 3/29/11 3.7.12-1 253/248 10/30/09 3.7.12-2 263/259 3/29/11 Catawba Units 1 and 2 Page 8 5/4/17

3.7.13-1 3~7.13-2 3.7.14-1 3.7.15-1 198/191 263/259 263/259 263/259 4/23/02 3/29/11 3/29/11 3/29/11 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 3.7.17-1 263/259 3/29/11 3.8.1-1 253/248 10/30/09 3.8.1-2 173/165 9/30/98 3.8.1-3 253/248 10/30/09 3.8.1-4 173/165 9/30/98 3.8.1-5 263/259 3/29/11 3.8.1-6 263/259 3/29/11 3.8.1-7 263/259 3/29/11 3.8.1-8 263/259 3/29/11 3.8.1-9 263/259 3/29/11 3.8.1-10 263/259 3/29/11 3.8.1-11 263/259 3/29/11 3.8.1-12 263/259 3/29/11

. 3.8.1-13 263/259 3/29/11 3.8.1-14 263/259 3/29/11 3.8.1-15 263/259 3/29/11 3.8.2-1 173/165 9/30/98 3.8.2-2 207/201 7/29/03 3.8.2-3 173/165 9/30/98 3.8.3-1 175/167 1/15/99 3.8.3-2 263/259 3/29/11 3.8.3-3 263/259 3/29/11 3.8.4-1 173/165 9/30/98 3.8.4-2 263/259 3/29/11 3.8.4-3 263/259 3/29/11 3.8.4-4 263/259 3/29/11 3.8.4-5 262/258 12/20/10 Catawba Units 1 and 2 Page 9 5/4/17

3.8.5-1 3.8.5-2 3.8.6-1 3.8.6-2 173/165 207/201 253/248 253/248 9/30/98 7129103 10/30/09 10/30/09 3.8.6-3 253/248 10/30/09 3.8.6-4 263/259 3/29/11 3.8.6-5 223/218 4/27/05 3.8.7-1 173/165 9/30/98 3.8.7-2 263/259 3/29/11 3.8.8-1 173/165 9/30/98 3.8.8-2 263/259 3/29/11 3.8.9-1 173/165 9/30/98 3.8.9-2 173/165 9/30/98 3.8.9-3 263/259 3/29/11 3.8.10-1 207/201 7/29/03 3.8.10-2 263/259 3/29/11 3.9.1-1 263/259 3/29/11 3.9.2-1 215/209 6/21/04 3.9.2-2 263/259 3/29/11 3.9.3-1 227/222 9/30/05 3.9.3-2 263/259 3/29/11 3.9.4-1 207/201 7/29/03 3.9.4-2 282/278 4/26/17 3.9.5-1 207/201 7/29/03 3.9.5-2 282/278 4/26/17 3.9.6-1 263/259 3/29/11 3.9.7-1 263/259 3/29/11 4.0-1 284/280 6/21/16 4.0-2 233/229 9/27/06 5.1-1 273/269 2/12/15 5.2-1 273/269 2/12/15 5.2-2 273/269 2/12/15 5.2-3 Deleted 9/21/09 5.3-1 273/269 2/12/15 Catawba Units 1 and 2 Page 10 5/4/17

5.4-1 5.5-1 5.5-2 5.5-3 173/165 286/282 286/282 173/165 9/30/98 9/12/16 9/12/16 9/30/98 5.5-4 173/165 9/30/98 5.5-5 216/210 8/5/04 5.5-6 280/276 4/26/16 5.5-7 280/276 4/26/16 5.5-7a 280/276 4/26/16 5.5-8 280/276 4/26/16 5.5-9 280/276 4/26/16 5.5-10 280/276 4/26/16 5.5-11 280/276 4/26/16 5.5-12 280/276 4/26/16 5.5-13 280/276 4/26/16 5.5-14 280/276 4/26/16 5.5-15 280/276 4/26/16 5.5-16 280/276 4/26/16 5.5-17 280/276 4/26/16 5.5-18 280/276 4/26/16 5.5-19 280/276 4/26/16 5.6-1 222/217 3/31/05 5.6-2 253/248 10/30/09 5.6-3 222/217 3/31/05 5.6-4 284/280 6/21/16 5.6-5 275/271 4/14/15 5.6-6 280/276 4/26/16 5.7-1 273/269 2/12/15 5.7-2 173/165 9/30/98

Catawba Units 1 and 2 Page 11 5/4/17

ii iii BASES Revision 1 Revision 2 Revision 1 4/08/99 3/01/05 6/21/04 8 2.1.1-1 Revision 0 9/30/98 8 2.1.1-2 Revision 1 12/19/03 8 2.1.1-3 Revision 1 12/19/03

.82.1.2-1 Revision 0 9/30/98 8 2.1.2-2 Revision O 9/30/98 8 2.1.2-3 Revision 0 9/30/98 8 3.0-1 Revision 1 3/19/07 8 3.0-2 Revision 1 3/19/07 8 3.0-3 Revision 2 3/19/07 8 3.0-4 Revision 3 3/19/07 8 3.0-5 Revision 3 3/19/07 8 3.0-6 Revision 2 3/19/07 8 3.0-7 Revision 2 3/19/07 8 3.0-8 Revision 3 3/19/07 8 3.0-9 Revision 2 3/19/07 8 3.0-10 Revision 3 3/19/07 8 3.0-11 Revision 3 3/19/07 B 3.0-12 Revision 3 3/19/07 83.0-13 Revision 3 3/19/07 8 3.0-14 Revision 3 3/19/07 8 3.0-15 Revision 1 3/19/07 8 3.0-16 Revision 1 3/19/07 8 3.0-17 Revision O 3/19/07 8 3.0-18 Revision 0 3/19/07 83.0-19 Revision O 3/19/07 8 3 .1.1-1 thru Revision 3 5/05/11 8 3.1.1-6 8 3.1.2-1 thru Revision 2 5/05/11 8 3.1.2-5

Catawba Units 1 and 2 Page 12 5/4/17

8 3.1.3-1 thru 8 3.1.3-6 8 3.1.4-1 thru 8 3.1.4-9 Revision 2 Revision 1 4/14/15 5/05/11 8 3.1.5-1 thru Revision 2 5/05/11 8 3.1.5-4 8 3.1.6-1 thru Revision 1 5/05/11 8 3.1.6-6 8 3.1.7-1 Revision 0 9/30/98 8 3.1.7-2 Revision 2 1/08/04 8 3.1.7-3 Revision 2 1/08/04 8 3.1.7-4 Revision 2 1/08/04 8 3.1.7-5 Revision 2 1/08/04 8 3.1.7-6 Revision 2 1/08/04 8 3.1.8-1 thru Revision 2 5/05/11 8 3.1.8-6 8 3.2.1-1 thru Revision 4 5/05/11 8 3.2.1.-11 8 3.2.2-1 thru Revision 3 5/05/11 8 3.2.2-10 8 3.2.3-1 thru Revision 2 5/05/11 8 3.2.3-4 8 3.2.4-1 thru Revision 2 5/05/11 8 3.2.4-7 8 3.3.1-1 thru Revision 8 4/08/16 8.3.3.1-55 8 3.3.2-1 thru Revision 12 12/18/15 8 3.3.2-50 8 3.3.3-1 thru Revision 6 4/11/14 8.3.3.3-16 8 3.3.4-1 thru Revision 2 5/05/11 8 3.3.4-5 8 3.3.5-1 thru Revision 3 12/18/15 8 3.3.5-6 Catawba Units 1 and 2 Page 13 5/4/17

B 3.3.6-1 thru B 3.3.6-5 B 3.3.9-1 thru B 3.3.9-5 Revision 6 Revision 3 08/02/12 06/02/14 B 3.4.1-1 thru Revision 3 5/05/11 B 3.4.1-5 B 3.4.2-1 Revision O 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 thru Revision 2 5/05/11 B 3.4.3-6 B 3.4.4-1 thru Revision 2 5/05/11 B 3.4.4-3 B 3.4.5-1 thru Revision 3 5/05/11 B 3.4.5-6 B 3.4.6-1 thru Revision 5 4/26/17 B 3.4.6-6 B 3.4.7-1 thru Revision 7 4/26/17 B 3.4.7-7 B 3.4.8-1 thru Revision 4 4/26/17 B 3.4.8-4 B 3.4.9-1 thru Revision 3 08/02/12 B 3.4.9-5 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 2 10/30/09 B 3.4.11-1 thru Revision 4 5/05/11 B 3.4.11-7 B 3.4.12-1 thru Revision 5 8/19/15 B 3.4.12-14 B 3.4.13-1 thru Revision 7 3/15/12 B 3.4.13-7

Catawba Units 1 and 2 Page 14 5/4/17

B 3.4.14-1 thru B 3.4.14-6 B 3.4.15-1 thru B 3.4.15-10 Revision 3 Revision 6 5/05/11 5/05/11 B 3.4.16-1 thru Revision 4 10/23/12 B 3.4.16-5 B 3.4.17-1 thru Revision 2 5/05/11 B 3.4.17-3 B 3.4.18-1 thru Revision 2 4/26/16 B 3.4.18-8 B 3.5.1-1 thru Revision 4 4/26/17 B 3.5.1-8 B 3.5.2-1 thru Revision 4 4/26/17 B 3.5.2-11 B 3.5.3-1 thru Revision 2 4/26/17 B 3.5.3-3 B 3.5.4-1 thru Revision 5 4/11/14 B.3.5.4-5 B 3.5.5-1 thru Revision 1 5/05/11 B 3.5.5-4 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 7/31/01 B 3.6.2-1 thru Revision 2 5/05/11 B 3.6.2-8 B 3.6.3-1 thru Revision 5 9/12/16 B 3.6.3-14 B 3.6.4-1 thru Revision 2 5/05/11 B 3.6.4-4 B 3.6.5-1 thru Revision 3 07/27/13 B 3.6.5-4

Catawba Units 1 and 2 Page 15 5/4/17

B 3.6.6-1 thru B 3.6.6-8 B 3.6.8-1 thru B 3.6.8-5 Revision 7 Revision 3 4/26/17 5/05/11 B 3.6.9-1 thru Revision 6 5/05/11 B 3.6.9-5 B 3.6.10-1 thru Revision 2 5/05/11 B 3.6.10-6 B 3.6.11-1 thru Revision 5 5/05/11 B 3.6.11-6 B 3.6.12-1 thru Revision 5 5/05/11 83.6.12-11 B 3.6.13-1 thru B Revision 4 5/05/11 3.6.13-9 B 3.6.14-1 thru Revision 2 4/11/14 B 3.6.14-5 B 3.6.15-1 thru Revision 1 5/05/11 B 3.6.15-4 B 3.6.16-1 thru Revision 3 5/05111 B 3.6.16-4 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 O 9/30/98 B 3.6.17-4 Revision O 9/30/98 B 3.6.17-5 Revision 1 3/13/08 B 3.7.1-1 thru Revision 2 4/29/16 3.7.1-5 B 3.7.2-1 Revision 0 9/30/98 B 3.7.2-2 Revision O 9/30/98 B 3.7.2-3 Revision 2 6/23/10 B 3.7.2-4 Revision 1 9/08/08 B 3.7.2-5 Revision 3 10/30/09 B 3.7.3-1 Revision 0 9/30/98 B 3.7.3-2 Revision O 9/30/98 Catawba Units 1 and 2 Page 16 5/4/17

B 3.7.3-3 B 3.7.3-4 B 3.7.3-5 B 3.7.3-6 Revision O Revision 0 Revision 1 Revision 2 9/30/98 9/30/98 9/08/08 10/30/09 B 3.7.4-1 thru Revision 2 5/05/11 B 3.7.4-4 B 3.7.5-1 thru Revision 3 5/05/11 B 3.7.5-9 B 3.7.6-1 thru Revision 4 08/02/12 B 3.7.6-3 B 3.7.7-1 thru Revision 2 5/05/11 B 3.7.7-5 B 3.7.8-1 thru Revision 5 08/09/13 B 3.7.8-8 B 3.7.9-1 thru Revision 3 5/05/11 B 3.7.9-4 B 3.7.10-1 thru B Revision 10 10/24/11 3.7.10-9 B 3.7.11-1 thru Revision 3 10/24/11 B 3.7.11-4 B 3.7.12-1 thru Revision 6 1/09/13 B 3.7.12-7 B 3.7.13-1 thru Revision 4 5/05/11 B 3.7.13-5 B 3.7.14-1 thru Revision 2 5/05/11 B 3.7.14-3 B 3.7.15-1 thru Revision 2 5/05/11 B 3.7.15-4 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 9127106 B 3.7.16-4 Revision O 9/27/06 B 3.7.17-1 thru Revision 2 5/05/11 B 3.7.17-3 Catawba Units 1 and 2 Page 17 5/4/17

B 3.8.1-1 thru B.3.8.1-29 B 3.8.2-1 B 3.8.2-2 Revision 5 Revision O Revision 0 07/27/13 9/30/98 9/30/98 B 3.8.2-3 Revision O 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 thru Revision 4 5/05/11 B 3.8.3-8 B 3.-8.4-1 thru Revision 10 5/05/11 83.8.4.10 B 3.8.5-1 Revision O 9/30/98 B 3.8.5-2 Revision 2 7/29/03 B 3.8.5-3 Revision 1 7129103 B 3.8.6-1 thru Revision 4 5/05/11 B 3.8.6-7 B 3.8.7-1 thru Revision 3 5/05/11 B 3.8.7-4 B 3.8.8-1 thru Revision 3 5/05/11 B 3.8.8-4 B 3.8.9-1 thru Revision 2 5/05/11 B 3.8.9-10 B 3.8.10-1 thru Revision 3 5/05/11 B 3.8.10-4 B 3.9.1-1 thru Revision 3 5/05/11 B 3.9.1-4 B 3.9.2-1 thru Revision 6 3/21/17 B 3.9.2-3 B 3.9.3-1 thru Revision 4 5/05/11 B 3.9.3-5 B 3.9.4-1 thru Revision 5 4/26/17 B 3.9.4-6

Catawba Units 1 and 2 Page 18 5/4/17

B 3.9.5-1 thru B 3.9.5-5 B 3.9.6-1 thru B 3.9.6-3 Revision 4 Revision 2 4/26/17 5/05/11 B 3.9.7-1 thru Revision 1 5/05/11 B 3.9.7-3

Catawba Units 1 and 2 Page 19 5/4/17

RCS Loops-MODE 4 B 3.4.6

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.6 RCS Loops-MODE 4 BASES BACKGROUND In MODE 4, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

The reactor coolant is circulated through four RCS loops connected in parallel to the reactor vessel, each loop containing an SG, a reactor coolant pump (RCP), arid appropriate flow, pressure, level, and temperature instrumentation for control, protection, and indication. The RCPs circulate the coolant through the reactor vessel and SGs at a sufficient rate to ensure proper heat transfer and to prevent boric acid stratification.

In MODE 4, either RCPs or RHR loops can be used to provide forced circulation. The intent of this LCO is to provide forced flow from at least one RCP or one RHR loop for decay heat removal and transport. The

APPLICABLE flow provided by one RCP loop or RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that two paths be available to provide redundancy for decay heat removal.

In MODE 4, RCS circulation is considered in the determination of the SAFETY ANALYSES time available for mitigation of the accidental boron dilution event. The RCS and RHR loops provide this circulation.

RCS Loops-MODE 4 satisfy Criterion 4 of 10 CFR 50.36 (Ref. 1).

LCO The purpose of this LCO is to require that at least two loops be OPERABLE in MODE 4 and that one of these loops be in operation. The LCO allows the two loops that are required to be OPERABLE to consist of any combination of RCS loops and RHR loops. Any one loop in operation provides enough flow to remove the decay heat from the core with forced circulation. An additional loop is required to be OPERABLE to provide redundancy for heat removal.

Note 1 permits all RCPs or RHR pumps to be de-energized for::;; 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests that are designed to validate various accident analyses values. One of the tests

Catawba Units 1 and 2 B 3.4.6-1 Revision No. 5

RCS Loops - MODE 4 B 3.4.6 BASES LCO (continued) performed during the startup testing program is the validation of rod drop times during cold conditions, both with and without flow. The no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits the de-energizing of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values must be revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to perform the test, and operating experience has shown that boron-stratification is not a problem during this short period with no forced flow.

Utilization of Note 1 is permitted provided the following conditions are met along with any other conditions imposed by initial startup test procedures:

a. No operations are permitted that would dilute the RCS boron concentration with coolant with boron concentrations less than required to meet SOM of LCO 3.1.1 and maintain kett < 0.99, therefore maintaining an adequate margin to criticality. Boron reduction with coolant at boron concentrations less than required to

assure SDM and maintain kett < 0.99 is prohibited because a 1

uniform concentration distribution throughout the RCS cannot be b.

ensured when in natural circulation; and Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause, a natural circulation flow obstruction.

Note 2 requires that the secondary side water temperature of each SG be

~ 50°F above each of the RCS cold leg temperatures before the start of an RCP with any RCS cold leg temperature~ 210°F. This restraint is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG, which has the minimum water level specified in SR 3.4.6.2. The water level is maintained by an OPERABLE AFW train in accordance with LCO 3.7.5, "Auxiliary Feedwater System."

Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required. Management of gas voids is important to RHR System OPERABILITY.

Catawba Units 1 and 2 B 3.4.6-2 Revision No. 5

RCS Loops - MODE 4 B 3.4.6

BASES APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing.

One loop of either RCS or RHR provides sufficient circulation for these purposes. However, two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to meet single failure considerations.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.4.17, "RCS Loops-Test Exceptions";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS If only one RCS loop is OP.ERABLE and two RHR loops are inoperable,

redundancy for heat removal is lost. Action must be initiated to restore a second RCS or RHR loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

If only one RHR loop is OPERABLE and in operation and there are no RCS loops OPERABLE, an inoperable RCS or RHR loop must be restored to OPERABLE status to provide a redundant means for decay heat removal.

If the parameters that are outside the limits cannot be restored, the unit must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to MODE 5 is a conservative action with regard to decay heat removal. With only one RHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining RHR loop, it would be safer to initiate that loss from MODE 5 (::; 200°F) rather than MODE 4 (200 to < 350°F).

The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an orderly manner and without challenging plant systems .

Catawba Units 1 and 2 B 3.4.6-3 Revision No. 5

RCS Loops - MODE 4 B 3.4.6 BASES ACTIONS (continued)

C.1 and C.2 If no loop is OPERABLE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet SOM of LCO 3.1.1 and maintain kett < 0.99 must be suspended and action to restore one RCS or RHR loop to OPERABLE status and operation must be initiated. RCP seal injection flow is not considered to be an operation involving a reduction in RCS boron concentration. The' required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SOM of LCO 3.1.1 and maintain kett < 0.99 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SOM and kett requirements maintains acceptable margin to criticality. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This SR requires verification that one RCS or RHR loop is in operation.

Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Sun/eillance Frequency Control Program.

SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is

~ 12%. If the SG secondary side narrow range water level is< 12%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Catawba Units 1 and 2 B 3.4.6-4 Revision No. 5

RCS Loops - MODE 4 B 3.4.6

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.6.4 RHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required RHR loop(s) and may also prevent water hammer, pump cavitation, and pumping ofnoncondensible gas into the reactor vessel.

Selection of RHR System locations susceptible to gas a-ccumulation is based on a review of system design information, including piping and

instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump),

the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be

Catawba Units 1 and 2 B 3.4.6-5 Revision No. 5

RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE REQUIREMENTS (continued) verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

This SR is modified by a Note that states the SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 4. In a rapid shutdown, there may be insufficient time to verify all susceptible locations prior to entering MODE 4.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

/

REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

Catawba Units 1 and 2 B 3.4.6-6 Revision No. 5

RCS Loops-MODE 5, Loops Filled B 3.4.7

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.7 RCS Loops-MODE 5, Loops Filled BASES BACKGROUND In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and transfer this heat either to the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers. While the principal means for decay heat removal is via the RHR System, the SGs are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.

In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RHR loops connected to the RCS, each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication. One

RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification.

The number of loops in operation can vary to suit the operational needs.

The intent of this LCO is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal.

The LCO provides for redundant paths of decay heat removal capability.

The first path can be an RHR loop that must l:>e OPERABLE and in operation. The second path can be another OPERABLE RHR loop or maintaining two SGs with secondary side narrow range water levels ~

12% to provide an alternate method for decay heat removal.

APPLICABLE In MODE 5, RCS circulation is considered in the determination of the time SAFETY ANALYSES available for mitigation of the accidental boron dilution event. The RHR loops provide this circulation.

RCS Loops-MODE 5 (Loops Filled) satisfy Criterion 4 of 10 CFR 50.36 (Ref. 1) .

Catawba Units 1 and 2 B 3.4.7-1 Revision No. 7

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO The purpose of this LCO is to require that at least one of the RHR loops be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs with secondary side narrow range water level ; : : 12%. One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant undE;ir these conditions. An additional RHR loop is required to be OPERABLE to meet single failure considerations. However, if the standby RHR loop is not OPERABLE, an acceptable alternate method is two SGs with their secondary side narrow

range water levels;;:::: 12%. Should the operating RHR loop fail, the SGs could be used to remove the decay heat. NRC Information Notices 94-36 and 95-35 and Westinghouse Nuclear Safety Advisory Letter 94-13 address concerns with crediting of natural circulation alternate cooling as backup to the loss of an RHR pump. Specifically, when the RCS is depressurized, there is potential for gas accumulation in the high points in the SG tubes and steam voiding in the SG tubes if the RCS pressure boundary is not intact such that the RCS can pressurize to maintain the reactor coolant subcooled to prevent steam void formation. Gas or steam void formation in the RCS loops could impede natural circulation. The B and C RCS loops are not credited as an acceptable alternate method of heat removal when all four reactor coolant pumps (RCPs) are off due to the potential for gas accumulation at the top of the SG tubes. RHR flow results in higher flow of gas-saturated water to the hot legs associated with the RHR suction drop. Therefore, gas accumulation at the high points of RCS loops supplying suction to the RHR pumps is more likely to result in gas accumulation that could impede natural circulation.

"Loops Filled" is defined as follows in operating procedures:

For RCS loops to be considered filled, the RCS pressure boundary must be intact or able to be quickly restored by control room action.

The RCS must be capable of pressurizing to the Low Temperature Overpressure Protection (LTOP) setpoint of 400 psig. If the current outage requires a fill and vent of the RCS, a fill and vent has been procedurally completed. The RCS must also meet level conditions that ensure sufficient pressure at the top of the SG tubes to prevent steam void formation when the pressurizer is vented to atmosphere.

Note 1 permi.ts all RHR pumps to be de-energized ~ 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to permit tests designed to validate various accident analyses values. One of the tests performed during the startup testing program is the validation of rod drop times during cold conditions, both with and without flow. The no flow test may be performed in MODE 3, 4, or 5 and requires that the pumps be stopped for a short period of time. The Note permits de-energizing of the pumps in order to perform this test and validate the assumed analysis values. If changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values must be revalidated by conducting the test again. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months time period is adequate to perform Catawba Units 1 and 2 B 3.4.7-2 Revision No. 7

RCS Loops - MODE 5, Loops Filled B 3.4.7

BASES LCO (continued) the test, and operating experience has shown that boron stratification is not likely during this short period with no forced flow.

Utilization of Note 1 is permitted provided the following conditions are met, along with any other conditions imposed by initial startup test procedures:

a. No operations are permitted that would dilute the RCS boron concentration with coolant with boron concentrations less than required to meet SOM of LCO 3.1.1, therefore maintaining an adequate margin to criticality. Boron reduction with coolant at boron concentrations less than required to assure SOM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

Note 2 allows one RHR loop to be inoperable for a period of up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop

during the only time when such testing is safe and possible .

Note 3 requires that the secondary side water temperature of each SG be s 50°F above each of the RCS cold leg temperatures before the start of an RCP with an RCS cold leg temperature s 210°F. This restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.

Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.

An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger.

If not in its normal RHR alignment from the RCS hot leg and returning to the RCS cold legs, the required RHR loop is OPERABLE provided the system may be placed in service from. the control room, or may be placed in service in a short period of time by actions outside the control room and there are no restraints to placing the equipment in service. RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. SGs A or D can perform as OPERABLE heat sinks with adequate water level; SGs B or C can perform as OPERABLE heat sinks with adequate water level and at least one RCP in operation .

Catawba Units 1 and 2 B 3.4.7-3 Revision No. 7

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO (continued)

Management of gas voids is important to RHR System OPERABILITY.

APPLICABILITY In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side narrow range water level of at least two SGs is required to be~ 12%.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled";

LCO 3.4.17-"RCS Loops-Test Exceptions";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side narrow range water levels< 12%, redundancy for heat removal is lost.

Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat'removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet SOM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. RCP seal injection flow is not considered to be an operation involving a reduction in RCS boron concentration. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SOM of LCO 3.1.1 _is required to assure continued safe Catawba Units 1 and 2 B 3.4.7-4 Revision No. 7

RCS Loops - MODE 5, Loops Filled

.B 3.4.7

BASES ACTIONS (continued) operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SOM maintains acceptable margin to criticality. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification that the required loop is in operation.

Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow range water levels are ;;:: 12% ensures an alternate

decay heat removal method in the event that the second RHR loop is not OPERABLE. The A and D SGs are OPERABLE with .:::_ 12% narrow range level. The B and C SGs are OPERABLE if narrow range levels are.:::, 12%

and at least one RCP is in operation. If both RHR loops are OPERABLE, this Surveillance is not needed. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.7.3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the RHR pump.

If secondary side narrow range water level is ;;:: 12% in at least two SGs, this Surveillance is not needed. The A and D SGs are OPERABLE with .:::,

12% narrow range level. The B and C SGs are OPERABLE if narrow range levels are.:::, 12% and at least one RCP is in operation. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program .

Catawba Units 1 and 2 B 3.4.7-5 Revision No. 7

RCS, Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.7.4 RHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the required RHR loop(s) and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of RHR System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the

acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump),

the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical. for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

Catawba Units 1 and 2 B 3.4.7-6 Revision No. 7

RCS Loops - MODE 5, Loops Filled B 3.4.7

BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

2. Problem Investigation Process M-00-02276 .

Catawba Units 1 and 2 B 3.4.7-7 Revision No. 7

RCS Loops-MODE 5, Loops Not Filled B 3.4.8

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.8 RCS Loops-MODE 5, Loops Not Filled BASES BACKGROUND In MODE 5 with the RCS loops not filled, the primary function of the reactor coolant is the removal of decay heat generated in the fuel, and the transfer of this heat to the component cooling water via the residual heat removal (RHR) heat exchangers. The steam generators (SGs) are not available as a heat sink when the loops are not filled. The secondary function of the reactor coolant is to act as a carrier for the soluble neutron poison, boric acid.

In MODE 5 with loops not filled, only RHR pumps can be used for coolant circulation. The number of pumps in operation can vary to suit the operational needs. The intent of this LCO is to provide forced flow from at least one RHR pump for decay heat removal and transport and to require that two paths be available to provide redundancy for heat removal.

APPLICABLE In MODE 5, RCS circulation is considered in the determination of the time SAFETY ANALYSES time available for mitigation of the accidental boron dilution event. The RHR loops provide this circulation. The flow provided by one RHR loop is adequate for heat removal and for boron mixing.

RCS loops in MODE 5 (loops not filled) satisfy Criterion 4 of 10 CFR 50.36 (Ref. 1).

LCO The purpose of this LCO is to require that at least two RHR loops be OPERABLE and one of these loops be in operation. An OPERABLE loop is one that has the capability of transferring heat from the reactor coolant at a controlled rate. Heat cannot be removed via the RHR System unless forced flow is used. A minimum of one running RHR pump meets the LCO requirement for one loop in operation. An additional RHR loop is required to be OPERABLE to meet single failure considerations.

Note 1 permits all RHR pumps to be de-energized for:::; 15 minutes when switching from one loop to another. The circumstances for stopping both RHR pumps are to be limited to situations when the outage time is short and core outlet temperature is maintained at least 10°F below saturation temperature. The Note prohibits boron dilution with coolant at boron concentrations less than required to assure SOM of LCO 3.1.1 is maintained or draining operations when RHR forced flow is stopped .

Catawba Units 1 and 2 B 3.4.8-1 Revision No. 4

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO (continued)

Note 2 allows. one RHR loop to be inoperable for a period of:::; 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , provided that the other loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when these tests are safe and possible.

An OPERABLE RHR loop is comprised of an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. If not in its normal RHR alignment from the RCS hot leg and returning to the RCS cold legs, the required RHR loop is OPERABLE provided the system may be placed in service from the control room, or may be placed in service in a short period of time by actions outside the control room and there are no restraints to placing the equipment in service. Management of gas voids is important to RHR System OPERABILITY.

APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the RHR System.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops-MODES 1 and 2";

LCO 3.4.5, "RCS Loops-MODE 3";

LCO 3.4.6, "RCS Loops-MODE 4";

LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled";

LCO 3.4.17, "RCS Loops-Test Exceptions";

LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level" (MODE 6); and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level" (MODE 6).

ACTIONS If only one RHR loop is OPERABLE and in operation, redundancy for RHR is lost. Action must be initiated to restore a second loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

B.1 and B.2 If no required RHR loops are OPERABLE or in operation, except during conditions permitted by Note 1, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet Catawba Units 1 and 2 B 3.4.8-2 Revision No. 4

RCS Loops - MODE 5, Loops Not Filled B 3.4.8

BASES ACTIONS (continued)

SDM of LCO 3.1.1 must be suspended and action must be initiated immediately to restore an RHR loop to OPERABLE status and operation.

RCP seal injection flow is not considered to be an operation involving a reduction in RCS boron concentration. The required margin to criticality must not be reduced in this type of operation. Suspending the introduction of coolant into the RCS of coolant with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however, coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to criticality. The immediate Completion Time reflects the importance of maintaining operation for heat removal.

The action to restore must continue until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification that one loop is in operation. Verification includes flow rate, temperature, or pump status monitoring, which help

ensure that forced flow is providing heat removal. The Surveillance Frequency is based on' operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.8.2 Verification that the required number of pumps are OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pumps. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.8.3 RHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR loops and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Catawba Units 1 and 2 B 3.4.8-3 Revision No. 4

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE REQUIREMENTS (continued)

Selection of RHR System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump),

the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is .controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES 1. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

Catawba Units 1 and 2 B 3.4.8-4 Revision No. 4

Accumulators B 3.5.1

B 3.5 EMERGENCY CORE COOLING SYSTEMS (EGGS)

B 3.5.1 Accumulators BASES BACKGROUND The functions of the EGGS accumulators are to supply water to the reactor vessel during the blowdown phase of a loss of coolant accident

. (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide Reactor Coolant System (RCS) makeup for a small break LOCA.

The blowdown phase. of a large break LOCA is the initial period of the transient during which the RCS departs from equilibrium conditions, and heat from fission product decay, hot internals, and the vessel continues to be transferred to the reactor coolant. The blowdown phase of the transient ends when the RCS pressure falls to a value approaching that of the containment atmosphere.

In the refill phase of a LOCA, which immediately follows the blowdown phase, reactor coolant inventory has vacated the core through steam flashing and ejection out through the break. The core is essentially in adiabatic heatup. The balance of accumulator inventory is then available to help fill voids in the lower plenum and reactor vessel downcomer so as to estabfish a recovery level at the bottom of the core and ongoing reflood of the core with the addition of safety injection (SI) water.

The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. The accumulators are passive components, since no operator or control actions are required in order for them to perform their function. Internal accumulator tank pressure is sufficient to discharge the accumulator contents to the RCS, if RCS pressure decreases below the accumulator pressure.

Each ac~umulator is piped into an RCS.cold leg via an accumulator line and is isolated from the RCS by a motor operated isolation valve and two check valves in series. The motor operated isolation valves are interlocked by P-11 with the pressurizer pressure measurement channels to ensure that the valves will automatically open as RCS pressure increases to above the. permissive circuit P-11 setpoint.

This interlock also prevents inadvertent closure of the valves during normal operation prior to an accident. The valves will automatically open, however, as a result of an SI signal. The isolation valves between the accumulators and the Reactor Coolant System are required to be open

Catawba Units 1 and 2 B 3.5.1-1 Revision No. 4

Accumulators B 3.5.1 BASES BACKGROUND (continued) and power removed during unit operation. In that the subject valves are normally open and do not serve as an active device during a LOCA, the requirements of the Institute of Electrical and Electronic Engineers (IEEE)

Standard 279-1971 (Ref. 1) is not applicable in this situation. Therefore, the subject valve control circuit is not designed to this standard.

The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.

APPLICABLE The accumulators are assumed OPERABLE in both the large and SAFETY ANALYSES small break LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is required by regulations and conservatively imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.

The limiting large break LOCA is a double ended guillotine break at the discharge of the reactor coolant pump. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.

As a conservative estimate, no credit is taken for ECCS pump flow until an effective delay has elapsed. This delay accounts for the diesels starting, the valves opening, and the pumps being loaded and delivering full flow. The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.

Catawba Units 1 and 2 B3.5.1-2 Revision No. 4

Accumulators B 3.5.1

BASES APPLICABLE SAFETY ANALYSES (continued)

The worst case small break LOCA analyses also assume a time delay before pumped flow reaches the core. For the larger range of small breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators, safety injection pumps, and centrifugal charging pumps all play a part in terminating the rise in clad temperature. As break size continues to decrease, the role of the accumulators continues to decrease until they are not required and the centrifugal charging pumps become solely responsible for terminating the temperature increase.

This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46 (Ref. 3) will be met following a small break LOCA and there is a high level of probability that the criteria are met following a large break LOCA:

a. Maximum fuel element cladding temperature is :s; 2200°F;

b. Maximum cladding oxidation is :s; 0.17 times the total cladding thickness before oxidation;

c.

d.

Maximum hydrogen generation from a zirconium water reaction is :s; 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and Core is maintained in a coolable geometry.

Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute directly to the long term cooling requirements of 10 CFR 50.46. However, the boron content of the accumulator water helps to maintain the reactor core subcritical after reflood, thereby eliminating fission heat as an energy source for which cooling must be provided.

For both the large and small break LOCA analyses, a nominal contained accumulator water volume is used. The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. The large and small break LOCA analyses are performed with accumulator volumes that are consistent with the LOCA evaluation models. To allow for operating margin, values of+/- 30 ft3 are specified.

The minimum boron concentration setpoirit is used in the post LOCA subcriticality verification during the injection phase. For each reload

Catawba Units 1 and 2 B 3.5.1-3 Revision No. 4

Accumulators B 3.5.1 BASES APPLICABLE SAFETY ANALYSES (continued) cycle, the all rods out (ARO) critical boron concentration is verified to be less than the minimum allowed cold leg accumulator boron concentration.

The minimum boron concentration setpoint is also used in the post LOCA sump boron concentration calculation. The calculation is performed to assure reactor subcriticality in a post LOCA environment with all rods in, minus the highest worth rod out (ARI N-1). Of particular interest is the large cold leg break LOCA, since boron accumulation in the core will be maximized during the cold leg recirculation phase due to core boiling.

The accumulation of boron in the core prevents the boron from returning to the sump, which leads to a boron diluted sump condition which may cause the core to become re-critical when switching over to hot leg recirculation. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH. In particular, the equilibrium sump pH should be at least 7.5 following the design basis LOCA.

The large and small break LOCA analyses are performed with accumulator pressures that are consistent with the LOCA evaluation

models. To allow for operating margin and accumulator design limits, a range from 585 psig to 678 psig is specified. The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity.

The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Ref. 4).

The accumulators satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).

LCO The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 3) could be violated.

For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 1000 psig, and the limits established in the SRs for contained volume, boron concentration, and Catawba Units 1 and 2 B 3.5.1-4 Revision No. 4

Accumulators B 3.5.1

BASES LCO (continued) nitrogen cover pressure must be met. Additionally, the nitrogen and liquid volumes between accumulators must be physically separate.

APPLI CASI LITY In MODES 1 and 2, and in MODE 3 with RCS pressure> 1000 psig, the accumulator OPERABILITY requirements are based on full power operation. Although cooling requirements decrease as power decreases, the accumulat0rs are still required to provide core cooling as long as elevated RCS pressures and temperatures exist.

This LCO is only applicable at pressures> 1000 psig. At pressures

~ 1000 psig, the rate of RCS blowdown is such that the ECCS pumps can provide adequate injection to ensure that peak claa temperature remains below the 10 CFR 50.46 (Ref. 3) limit of 2200°F for small break LOCAs and there is a high level of probability that the peak cladding temperature does not exceed 2200°F for large break LOCAs.

In MODE 3, with RCS pressure~ 1000 psig, and in MODES 4, 5, and 6, the accumulator motor operated isolation valves are allowed to be closed to isolate the accumulators from the RCS. This allows RCS cooldown

ACTIONS and depressurization without discharging the accumulators into the RCS or requiring depressurization of the accumulators.

If the boron concentration of one accumulator is not within limits, it must be returned to within the limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, ability to maintain subcriticality or minimum boron precipitation time may be reduced. The boron in the accumulators contributes to the assumption that the combined ECCS water in the partially recovered core during the early reflooding phase of a large break LOCA is sufficient to keep that portion of the core subcritical. One accumulator below the minimum boron concentration limit, however, will have no effect on available ECCS water and an insignificant effect on core subcriticality during reflood.

Boiling of ECCS water in the core during reflood concentrates boron in the saturated liquid that remains in the core. In addition, current analysis techniques demonstrate that the accumulators do not discharge following a large main steam line break for the plant. Even if they do discharge, their impact is minor and not a design limiting event. Thus, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to return the boron concentration to within limits .

Catawba Units 1 and 2 B 3.5.1-5 Revision No. 4

Accumulators B 3.5.1 BASES ACTIONS (continued)

If one accumulator is inoperable for a reason other than boron concentration, the accumulator must be returned to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In this Condition, the required contents of three accumulators cannot be assumed to reach the core during a LOCA. Due to the severity, of the consequences should a LOCA occur in these conditions, the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time to open the valve, remove power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions.

The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-15049-A, Rev. 1 (Ref. 6).

C.1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in

which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and RCS pressure reduced to ::;

1000 psig 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 plant conditions from full power conditions in an orderly manner and without challenging plant systems.

If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.

Catawba Units 1 and 2 B 3.5.1-6 Revision No. 4

Accumulators B 3.5.1

BASES SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator valve should be verified to be fully open. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power* removed, a closed valve could result in not meeting accident analyses assumptions. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.5.1.2 and SR 3.5.1.3 Borated water volume and nitrogen cover pressure are verified for each accumulator. This is typically performed using the installed control room indication. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator since the static design of the accumulators limits the ways in which the concentration can be changed. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. Sampling the affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a 75 gallon increase will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 7) .

Catawba Units 1 and 2 B 3.5.1-7 Revision No. 4

Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.1.5

Verification that power is removed from each accumulator isolation valve operators for Nl54A, Nl65B, Nl76A, and Nl88B when the RCS pressure is

> 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is s; 1000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns. Even with power supplied to the valves, inadvertent closure is prevented by the RCS pressure interlock associated with the valves.

Should closure of a valve occur in spite of the interlock, the SI signal REFERENCES provided to the valves would open a closed valve in the event of a LOCA.

1.

2.

IEEE Standard 279-1971.

UFSAR, Chapter 6.

3. 10 CFR 50.46.

4. DPC-NE-3004.

5. 10 CFR 50.36, Technical Specification, (c)(2)(ii).

6. WCAP-15049-A, Rev. 1, April 1999.

7. NUREG-1366, February 1990.

Catawba Units 1 and 2 B 3.5.1-8 Revision No. 4

ECCS-Operating B 3.5.2

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.2 ECCS-Operating BASES BACKGROUND The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

a. .Loss of coolant accident (LOCA), coolant leakage greater than the capability of the normal charging system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steam or feedwater release; and

d. Steam generator tube rupture (SGTR).

The addition of negative reactivity is designed primarily for the loss of secondary coolant accident where primary cooldown could add enough positive reactivity to achieve criticality and return to significant power.

There are three phases of ECCS operation: injection, cold leg recirculation, and hot leg recirculation. In the injection phase, water is taken from the refueling water storage tank (RWST} and injected into the Reactor Coolant System (RCS) through the cold legs. When sufficient water is removed from the RWST to ensure that enough boron has been added to maintain the reactor subcritical and the containment sumps have enough water to supply the required net positive suction head to the ECCS pumps, suction is switched to the containment sump for cold leg recirculation. When the core decay heat has decreased to a level low enough to be successfully removed without direct RHR pump injection flow, the RHR cold leg injection path is.realigned to discharge to the auxiliary containment spray header. After approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , part of the ECCS flow is shifted to the hot leg recirculation phase to provide a backflush which, for a cold leg break, would reduce the boiling in the top of the core and prevent excessive boron concentration.

The ECCS consists of three separate subsystems: centrifugal charging (high head), safety injection (SI) (intermediate head), and residual heat

.removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains. The ECCS accumulators and the RWST are also part of the ECCS, but are not considered part of an ECCS flow path as described by this LCO .

Catawba Units 1 and 2 B 3.5.2-1 Revision No. 4

ECCS - Operating B 3.5.2 BASES BACKGROUND (continued)

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the RWST can be injected into the RCS following the accidents described in this LCO. The major components of each subsystem are the centrifugal charging pumps, the RHR pumps, heat exchangers, and the SI pumps. Each of the three subsystems consists of two 100% capacity trains that are interconnected and redundant such that either train is capable of supplying 100% of the flow required to mitigate the accident consequences. This interconnecting and redundant subsystem design provides the operators with the ability to utilize components from opposite trains to achieve the required 100%

flow to the core.

During the injection phase of LOCA recovery, a suction header supplies water from the RWST to the ECCS pumps. Mostly separate piping supplies each subsystem and each train within the subsystem. The discharge from the centrifugal charging pumps combines, then divides again into four supply lines, each of which feeds the injection line to one RCS cold leg. The discharge from the SI and RHR pumps divides and feeds an injection line to each of the RCS cold legs. Throttle valves in the SI lines are set to balance the flow to the RCS. This balance ensures sufficient flow to the core to meet the analysis assumptions following a LOCA in one of the RCS cold legs. The flow split from the RHR lines

cannot be adjusted. Although much of the two ECCS trains are composed of completely separate piping, certain areas are shared between trains. The most important of these are 1) where both trains flow through a single physical pipe, and 2) at the injection connections to the RCS cold legs. Since each train must supply sufficient flow to the RCS to be considered 100% capacity, credit is taken in the safety analyses for flow to three intact cold legs. Any configuration which, when combined with a single active failure, prevents the flow from either ECCS pump in a given train from reaching all four cold legs injection points on that train is unanalyzed and might render both trains of that ECCS subsystem inoperable.

For LOCAs that are too small to depressurize the RCS below the shutoff head of the SI pumps, the centrifugal charging pumps supply water until the RCS pressure decreases below the SI pump shutoff head. During this period, the steam generators are used to provide part of the core cooling function.

During the recirculation phase of LOCA recovery, RHR pump suction is transferred to the containment sump. The RHR pumps then supply the other ECCS pumps. Initially, recirculation is through the same paths as the injection phase. Subsequently, for large LOCAs, the recirculation phase includes injection into both the hot and cold legs.

Catawba Units 1 and 2 B 3.5.2-2 Revision No. 4

ECCS - Operating B 3.5.2

BASES BACKGROUND (continued)

The high and intermediate head subsystems of the ECCS also functions to supply borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB). The limiting design conditions occur when the moderator temperature coefficient is highly negative, such as at the end of each cycle.

During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," for the basis of these requirements.

The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is* accomplished in a programmed time sequence. If offsite power is available, the safeguard loads start immediately in the programmed sequence. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs). Safeguard loads are then actuated in the programmed time sequence. The time delay a~sociated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a safety injection actuation .

The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5.1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST)," provide the cooling water necessary to meet GDC 35 (Ref. 1).

APPLICABLE The LCO helps to ensure that the following acceptance criteria for the SAFETY ANALYSES ECCS, established by 10 CFR 50.46 (Ref. 2), will be met following a small break LOCA and there is a high level of probability that the criteria are met following a large break LOCA:

a. . Maximum fuel element cladding temperature is :,;; 2200°F;

b. Maximum Cladding oxidation is :,;; 0.17 times the total cladding thickness before oxidation;

c. Maximum hydrogen generation from a zirconium water reaction is

,;; 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;

Catawba Units 1 and 2 B 3.5.2-3 Revision No. 4

ECCS - Operating B 3.5.2 BASES APPLICABLE SAFETY ANALYSES (continued)

d. Core is maintained in a coolable geometry; and

e. Adequate long term core cooling capability is maintained.

The LCO also limits the potential for a post trip return to power following an MSLB event and ensures that containment pressure and temperature limits are met.

Each ECCS subsystem is taken credit for in a large break LOCA event at full power (Refs. 3 and 4). This event has the greatest potential to challenge the limits on runout flow set by the manufacturer of the ECCS pumps. It also sets the maximum response time for their actuation.

Direct flow from the centrifugal charging pumps and SI pumps is credited in a small break LOCA event. The RHR pumps are also credited, for larger small break LOCAs, as the means of supplying suction to these higher head ECCS pumps after the switch to ~ump recirculation. This event establishes the flow and discharge head at the design point for the centrifugal charging pumps. The MSLB analysis also credits the SI and centrifugal charging pumps. Although some ECCS flow is necessary to mitigate a SGTR event, a single failure disabling one ECCS train is not the limiting single failure for this transient. The SGTR analysis primary to secondary break flow is increased by the availability of both centrifugal

charging and SI trains. Therefore, the SGTR analysis is penalized by assuming both ECCS trains are operable as required by the LCO. The OPERABILITY requirements for the ECCS are based on the following LOCA analysis assumptions:

a. A large break LOCA event, with loss of offsite power and a single failure disabling one ECCS train; and

b. A small break LOCA event, with *a loss of offsite power and a single failure disabling one ECCS train.

During the blowdown stage of a LOCA, the RCS depressurizes as primary coolant is ejected through the break into the containment. The nuclear reaction is terminated either by moderator voiding during large breaks or control rod insertion for small breaks. Following depressurization, emergency cooling water is injected into the cold legs, flows into the downcomer, fills the lower plenum, and refloods the core.

The effects on containment mass and energy releases are accounted for in appropriate analyses (Ref. 3). The LCO ensures that an ECCS train will deliver sufficient water to match boiloff rates soon enough to minimize the consequences of the core being uncovered following a large LOCA.

Catawba Units 1 and 2 B 3.5.2-4 Revision No. 4

ECCS - Operating B 3.5.2

BASES APPLICABLE SAFETY ANALYSES (continued)

It also ensures that the centrifugal charging and SI pumps will deliver sufficient water and boron during a small LOCA to maintain core subcriticality. For smaller LOCAs, the centrifugal charging pump delivers sufficient fluid to maintain RCS inventory. For a small break LOCA, the steam generators continue to serve as the heat sink, providing part of the required core cooling.

The ECCS trains satisfy Criterion 3 of 10 CFR 50.36 (Ref. 5).

LCO In MODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that sufficient ECCS flow is available, assuming a single failure affecting either train. Additionally, individual components within the ECCS trains may be called upon to mitigate the consequences of other transients and accidents.

In MODES 1, 2, and 3, an ECCS train consists of a centrifugal charging subsystem, an SI subsystem, and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an SI signal and automatically transferring suction to the containment sump .

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the EGGS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to supply its flow to the RCS hot and cold legs. The flow path for each train must maintain its designed independence to ensure that no single failure can disable both ECCS trains. Management of gas voids is important to ECCS OPERABILITY.

APPLICABILITY In MODES 1, 2, and 3, the ECCS OPERABILITY requirements for the limiting Design Basis Accident, a large break LOCA, are based on full power operation. Although reduced power would not require the same level of performance, the accident analysis does not provide for reduced cooling requirements in the lower MODES. The centrifugal charging pump performance is based on a small break LOCA, which establishes the pump performance curve and has less dependence on power. The SI pump performance requirements are based on a small break LOCA.

For both of these types of pumps, the large break LOCA analysis depends only on the flow value at containment pressure, not on the shape of the flow versus pressure curve at higher pressures. MODE 2

Catawba Units 1 and 2 B 3.5.2-5 Revision No. 4

ECCS - Operating B 3.5.2 BASES APPLICABILITY (continued) and MODE 3 requirements are bounded by the MODE 1 analysis.

This LCO is only applicable in MODE 3 and above. Below MODE 3, the SI signal setpoint is manually bypassed by operator control, and system functional requirements are relaxed as described in LCO 3.5.3, "ECCS-Shutdown."

As indicated in the Note, the flow path may be isolated for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in MODE 3, under controlled conditions, to perform pressure isolation valve testing per SR 3.4.14.1. The flow path is readily restorable from the control room.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Residual Heat Removal *(RHR) and Coolant Circulation-High Water Level," and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level."

ACTIONS With one or more trains inoperable and at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the inoperable components must be returned to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on an NRC reliability evaluation (Ref. 6) and is a reasonable time for repair of many ECCS components.

An ECCS train is inoperable if it is not capable of delivering design flow to the RCS. Individual components are inoperable if they are not capable of performing their design function or supporting systems are not available.

The LCO requires the OPERABILITY of a number of independent subsystems. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does not render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of this Condition is to maintain a combination of equipment such that 100% of the ECCS flow equivalent to a single OPERABLE ECCS train remains available. This allows increased*flexibility in plant operations under circumstances when components in opposite trains are inoperable.

Catawba Units 1 and 2 B 3.5.2-6 Revision No. 4

ECCS - Operating B 3.5.2

BASES ACTIONS (continued)

An event accompanied by a loss of offsite power and the failure of an EOG can disable one ECCS train until power is restored. A reliability analysis (Ref. 6) has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Reference 7 describes situations in which one component, such as an RHR crossover valve, can disable both ECCS trains. With one or more component(s) inoperable such that 100% of the flow equivalent to a single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis~ Therefore, LCO 3.0.3 must be immediately entered.

B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 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 plant conditions from full power

conqitions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves using the power disconnect switches in the correct position ensures that they cannot change position as a result of an active failure or be inadvertently misaligned, These valves are of the type, described in Reference 7, that can disable th~ function of both ECCS trains and invalidate the accident analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing,

Catawba Units 1 and 2 B 3.5.2-7 Revision No. 4

ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

The Surveillance is modified by a Note which exempts system vent flow paths opened under administrative control. The administrative control should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication with the operators in the control room. This individual will have a method to rapidly close the system vent flow path if directed.

SR 3.5.2.3 ECCS piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and

accumulation is necessary for proper operation of the ECCS and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of ECCS locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The ECCS is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the ECCS is not rendered inoperable by the accumulated Catawba Units 1 and 2 B 3.5.2-8 Revision No. 4

ECCS - Operating B 3.5.2

BASES SURVEILLANCE REQUIREMENTS (continued) gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

ECCS locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety.

For these locations alte~native methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location.

Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

SR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis.

SRs are specified in the lnservice Testing Program, the ASME Code.

The ASME Code provides the activities and Frequencies necessary to satisfy the requirements .

Catawba Units 1 and 2 B 3.5.2-9 Revision No. 4

ECCS - Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.5.2.5 and SR 3.5.2.6 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI and Containment Sump Recirculation signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required f9r valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.7 The position of throttle valves in the flow path on an SI signal is necessary for proper ECCS performance. These valves have mechanical locks to ensure proper positioning for restricted flow to a ruptured cold leg, ensuring that-the other cold legs receive at least the required minimum flow. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.5.2.8 Periodic inspections of the containment sump suction inlet ensure that it is unrestricted and stays in proper operating condition. The Surveillance Frequency is based on operating experience, equipment reliability, and

plant risk and is controlled under the Surveillance Frequency Control Program.

Upon completion of the ECCS sump strainer assembly modifications during outage 2EOC15 for Unit 2 and 1EOC17 for Unit 1, the following SR Bases will apply:

Periodic inspections of the ECCS containment sump strainer assembly (consisting of modular tophats, grating, plenums, and waterboxes) ensure it is unrestricted and remains in proper operating condition.

Catawba Units 1 and 2 B 3.5.2-10 Revision No. 4

ECCS - Operating B 3.5.2

BASES SURVEILLANCE REQUIREMENTS (continued)

Inspections will consist of a visual examination of the exterior surfaces of the strainer assembly for any evidence of debris, structural distress or abnormal corrosion. The intent of this surveillance is to ensure the absence of any condition which could adversely affect strainer functionality'. Surveillance performance does not require removal of a.ny tophat modules or grating, but the strainer exteriors shall be visually inspected. This surveillance is not a commitment to inspect 100% of the surface area of all tophats, but a sufficiently detailed inspection of exterior strainer surfaces is required to establish a high confidence that no adverse conditions are present. The scope of inspection necessary to provide high confidence includes 100% of the strainer areas that can be accessed and inspected using normal means and tools (i.e., flashlight, extendable mirror, hand held digital camera) without disassembly, and that difficult to access areas will be inspected to the extent possible using these same means.

Any damage detected in the strainer assembly inspection will result in an expansion of the scope of the inspection to include other areas of potential damage. Inspection scope should be expanded, as needed, for degradation of strainer components identified during this inspection that were not considered readily accessible during the inspector's initial evaluation .

REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.

2. 10 CFR 50.46.

3. UFSAR, Section 6.2.1.

4. UFSAR, Chapter 15.

5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

6. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

7. IE Information Notice No. 87-01 .

Catawba Units 1 and 2 B 3.5.2-11 Revision No. 4

ECCS-Shutdown B 3.5.3

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.3 ECCS-Shutdown BASES BACKGROUND The Background section for Bases 3.5.2, "ECCS-Operating," is applicable to these Bases, with the following modifications.

In MODE 4, the required ECCS train consists of two separate subsystems: centrifugal charging (high head) and residual heat removal (RHR) (low head).

The ECCS flow paths consist of piping, valves, heat exchangers, and pumps such that water from the refueling water storage tank (RWST) can be injected into the Reactor Coolant System (RCS) following the accidents described in Bases 3.5.2.

APPLICABLE The Applicable Safety Analyses section of Bases 3.5.2 also applies SAFETY ANALYSES to this Bases section.

Due to the stable conditions associated with operation in MODE 4 and the reduced probability of occurrence of a Design Basis Accident (OBA),

the ECCS operational requirements are reduced. It is understood in these reductions that certain automatic safety injection (SI) actuation is not available. In this MODE, sufficient time exists for manual actuation of the required ECCS to mitigate the consequences of a OBA.

Only one train of ECCS is required for MODE 4. This requirement dictates that single failures are not considered during this MODE of operation. The ECCS trains satisfy Criterion 3 of 10 CFR 50.36.

LCO In MODE 4, one of the two independent (and redundant) ECCS trains is required to be OPERABLE to ensure that sufficient ECCS flow is available to the core following a OBA.

In MODE 4, an ECCS train consists of a centrifugal charging subsystem and an RHR subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST and transferring suction to the containment sump.

During an event requiring ECCS actuation, a flow path is required to provide an abundant supply of water from the RWST to the RCS via the

Catawba Units 1 and 2 B 3.5.3-1 Revision No. 2

ECCS - Shutdown B 3.5.3 BASES LCO (continued)

ECCS pumps and their respective supply headers to each of the four cold leg injection nozzles. In the long term, this flow path may be switched to take its supply from the containment sump and to deliver its flow to the RCS hot and cold legs. Management of gas voids is important to ECCS OPERABILITY.

APPLICABILITY In MODES 1, 2, and 3, the OPERABILITY requirements for ECCS are covered by LCO 3.5.2.

In MODE 4 with RCS temperature below 350°F, one OPERABLE ECCS train is acceptable without single failure consideration, on the basis of the stable reactivity of the reactor and the limited core cooling requirements.

In MODES 5 and 6, plant conditions are such that the probability of an event requiring ECCS injection is extremely low. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, "RCS Loops-MODE 5, Loops Filled," and LCO 3.4.8, "RCS Loops-MODE 5, Loops Not Filled." MODE 6 core cooling requirements are addressed by LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," and.LCO 3.9.5, "Residual Heat Removal (RHR) and

Coolant Circulation-Low Water Level."

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable ECCS centrifugal charging subsystem when entering MODE 4. There is an increased risk associated with entering MODE 4 from MODE 5 with an inoperable ECCS centrifugal charging subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

With no ECCS RHR subsystem OPERABLE, the plant is not prepared to respond to a loss of coolant accident or to continue a cooldown using the RHR pumps and heat exchangers. The Completion Time of immediately to initiate actions that would restore at least one ECCS RHR subsystem to OPERABLE status ensures that prompt action is taken to restore the required cooling capacity. Normally, in MODE 4, reactor decay heat is removed from the RCS by an RHR loop. If no RHR loop is OPERABLE Catawba Units 1 and 2 B 3.5.3-2 Revision No. 2

ECCS - Shutdown B3.5.3

BASES ACTIONS (continued) for this function, reactor decay heat must be removed by some alternate method, such as use of the steam generators. The alternate means of heat removal must continue until the inoperable RHR loop components can be restored to operation so that decay heat removal is continuous.

With both RHR pumps and heat exchangers inoperable, it would be unwise to require the plant to go to MODE 5, where the only available heat removal system is the RHR. Therefore, the appropriate action is to initiate measures to restore one ECCS RHR subsystem and to continue the actions until the subsystem is restored to OPERABLE status.

With no ECCS high head subsystem OPERABLE, due to the inoperability of the centrifugal charging pump or flow path from the RWST, the plant is not prepared to provide high pressure response to Design Basis Events requiring SI. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time to restore at least one ECCS high head subsystem to OPERABLE status ensures that prompt action is taken to provide the required cooling capacity or to initiate actions to place the plant in MODE 5, where an ECCS train is not required .

When the Required Actions of Condition B cannot be completed within the required Completion Time, a controlled shutdown should be initiated.

Twenty-four hours is a reasonable time, based on operating experience, to reach MODE 5 in an orderly manner and without challenging plant systems or operators.

SURVEILLANCE . SR 3.5.3.1 REQUIREMENTS The applicable Surveillance descriptions from Bases 3.5.2 apply. This SR is modified by a Note that allows an RHR train to be considered OPERABLE during alignment and operation for decay heat removal, if capable of being manually realigned (remote or local) to the ECCS mode of operation and not otherwise inoperable. This allows operation in the RHR mode during MODE 4, if necessary.

REFERENCES The applicable references from Bases 3.5.2 apply .

Catawba Units 1 and 2 B 3.5.3-3 Revision No. 2

Containment Isolation Valves B 3.6.3

B 3.6 CONTAINMENT SYSTEMS B 3.6.3 Containment Isolation Valves BASES BACKGROUND The containment isolation valves form part of the containment pressure

. boundary and provide a means for fluid penetrations not serving accident consequence limiting systems to be provided with two isolation barriers that are closed on a containment isolation signal. These isolation devices are either passi:ve or active (automatic). Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered passive devices. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. One of these barriers may be a closed system. These barriers (typically containment isolation valves) make up the Containment Isolation System .

Automatic isolation signals are produced during accident conditions.

Containment Phase "A" isolation occurs upon receipt of a safety injection signal. The Phase "A" isolation signal isolates nonessential process lines in order to minimize leakage of fission product radioactivity. Containment Phase "B" isolation occurs upon receipt of a containment pressure High-High signal and isolates the remaining process lines, except systems required for accident mitigation. The Containment Purge Ventilation and Containment Air Release and Addition valves also receive an isolation signal on a containment high radiation condition. As a result, the containment isolation valves (and blind flanges) help ensure that the containment atmosphere will be isolated from the environment in the event of a release of fission product radioactivity to the containment atmosphere as a result of a Design Basis Accident (OBA).

The OPERABILITY requirements for containment isolation valves help ensure that containment is isolated within the Time limits assumed in the safety analyses. Therefore, the OPERABILITY requirements provide assurance that the containment function assumed in the safety analyses will be maintained .

Catawba Units 1 and 2 B 3.6.3-1 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES BACKGROUND (continued)

Containment Purge Ventilation System The Containment Purge Ventilation System consists of the Containment Purge Supply and Exhaust Systems and the lncore Instrumentation Room Purge Supply and Exhaust Systems. These systems are used during refueling and post LOCA conditions and are not utilized during MODES 1 - 4. The penetration valves are sealed closed in MODES 1 -

4.

The Containment Purge Supply System includes one supply duct penetration through the Reactor Building wall into the annulus area.

There are four purge air supply penetrations through the containment vessel, two to the upper compartment and two to the lower compartment.

Two normally closed isolation valves at each penetration through the containment vessel provide containment isolation.

The Containment Purge Exhaust System includes 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.

The lncore Instrumentation Room Purge Supply System consists of one purge supply penetration through the Reactor Building wall and one through the containment vessel. Two normally closed isolation valves at the containment penetration provide containment isolation.

The lncore Instrumentation Room Purge Exhaust System consists of one purge exhaust penetration through the Reactor Building wall and one through the containment vessel. Two normally closed isolation valves at the penetration through the containment vessel provide containment isolation.

Containment Hydrogen Purge System The Containment Hydrogen Purge System co*nsists of a containment hydrogen purge inlet blower, which blows air from the Auxiliary Building through a 4 inch pipe into the upper compartment of the containment.

Another 4 inch pipe originating in the upper compartment of the containment purges air from the containment to the annulus. The penetration valves are sealed closed during MODES 1 - 4.

Catawba Units 1 and 2 B 3.6.3-2 Revision No. 5

Containment Isolation Valves B 3.6.3

BASES BACKGROUND (continued)

Containment Air Release and Addition System The Containment Air Release and Addition System is only used for controlling Containment pressure during normal unit operation. Isolation valves are located both inside and outside of the Containment on each containment penetration.

APPLICABLE The containment isolation valve LCO was derived from the assumptions SAFETY ANALYSES related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during major accidents. As part of the containment boundary, containment isolation valve OPERABILITY supports leak tightness of the containment. Therefore, the safety analyses of any event requiring isolation of containment is applicable to this LCO.

The DBAs that result in a release of radioactive material within containment are a loss of coolant accident (LOCA) and a rod ejection accident (Ref. 1). In the analyses for each of these accidents, it is assumed that containment isolation valves are either closed or function to

close within the required isolation time following event initiation. This ensures that potential paths to the environment through containment isolation valves (including containment purge valves) are minimized. The safety analyses assume that the containment purge supply and/or exhaust isolation valves for the lower compartment and the upper compartment, instrument room, and the Hydrogen Purge System are closed at event initiation. Since the Containment Purge Ventilation System and the Hydrogen Purge System isolation valves are sealed closed in MODES 1 - 4, they are not analyzed mechanistically in the dose calculations ..

The OBA analysis assumes that, within :::; 76 seconds after the accident, isolation of the containment is complete and leakage terminated except for the design leakage rate, La. The containment isolation total response time of:::; 76 seconds includes signal delay, diesel generator startup (for loss of offsite power), and containment isolation valve stroke times.

The single failure criterion required to be imposed in the conduct of plant safety analyses was considered in the original design of the containment purge valves. Two valves in series on each purge line provide assurance that both the supply and exhaust lines could be isolated even if a single failure occurred .

Catawba Units 1 and 2 B 3.6.3-3 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES APPLICABLE SAFETY ANALYSES (continued)

The containment purge and hydrogen purge valves may be unable to close in the environment following a LOCA. Therefore, each of the containment purge and hydrogen purge valves is required to remain sealed closed during MODES 1, 2, 3, and 4. The containment air release and addition valves may be opened during normal operation. In this case, the single failure criterion remains applicable to the containment air release and addition valves due to failure in the control circuit associated with each valve. The system valve design precludes a single failure from compromising the containment boundary as long as the system is operated in accordance with the subject LCO.

The containment isolation valves satisfy Criterion 3 of 10 CFR. 50.36 (Ref. 2).

LCO Containment isolation valves form a part of the containment boundary.

The containment isolation valves' safety function is related to minimizing the loss of reactor coolant inventory and establishing the containment boundary during a OBA.

The automatic power operated isolation valves are required to have

isolation times within limits and to actuate on an automatic isolation signal. The containment purge supply and exhaust isolation valves for the lower compartment, upper compartment, instrument room, and the Hydrogen Purge System must be maintained sealed closed. The valves covered by this LCO are listed along with their associated stroke times in the UFSAR (Ref. 3).

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves/devices are those listed in Reference 3.

Valves with resilient seals and reactor building bypass valves must meet additional leakage rate requirements. The other containment isolation valve leakage rates are addressed by LCO 3.6.1, "Containment," as Type C testing.

This LCO provides assurance that the containment isolation valves and purge valves will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the containment boundary during accidents.

Catawba Units 1 and 2 B 3.6.3-4 Revision No. 5

Containment Isolation Valves B 3.6.3

BASES APPLICABILITY In MODES 1, 2, 3, and 4, a OBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment isolation valves are not required to be OPERABLE in MODE 5. The requirements for containment isolation .valves during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations."

ACTIONS The ACTIONS are modified by a Note allowing penetration flow paths, except for containment purge supply and exhaust isolation valves for the lower and upper compartment, instrument room, and hydrogen purge penetration flow paths, to be unisolated intermittently under administrative controls. These administrative controls consist of stationing a dedicated operator at the valve controls, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated whe~ a need for containment isolation is indicated. For valve controls located in the control room, an operator may monitor containment isolation signal status rather than be stationed at the valve controls. Due to the size of the containment purge line penetration and the fact that those penetrations exhaust directly from the containment atmosphere to the environment, the penetration flow path containing

these valves may not be op'ened under administrative controls. A single purge valve in a penetration flow path may be opened to effect repairs to an inoperable valve,' as allowed by SR 3.6.3.1.

A second Note has been added to provide clarification that, for this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory acti9ns for each inoperable containment isolation valve. Complying with the Required Actions may allow for continued operation, and subsequent inoperable containment isolation valves are governed by subsequent Condition entry and application of associated Required Actions.

The ACTIONS are further modified by a third Note, which ensures appropriate remedial actions are taken, if necessary, if the affected systems are rendered inoperable by an inoperable containment isolation valve.

In the event the containment isolation valve leakage results in exceeding the overall containment leakage rate, Note 4 directs entry into the applicable Conditions and Required Actions of LCO 3.6.1 .

Catawba Units 1 and 2 B 3.6.3-5 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued)

A.1 and A.2 In the event one containment isolation valve in one or more penetration flow paths is inoperable except for purge valve or reactor building bypass leakage not within limit, the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure .. Isolation barriers that meet this criterion are a closed and de-activated automatic containment isolation valve, a closed manual valve, a blind flange, and a check valve inside containment with flow through the valve secured. For a penetration flow path isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available one to containment. Required Action A.1 must be completed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable, considering the time required to isolate the penetration and the relative importance of supporting containment OPERABILITY during MODES 1, 2, 3, and 4.

For affected penetration flow paths that cannot be restored to OPERABLE status within the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time and that have been isolated in accordance with Required Action A.1, the affected penetration flow paths must be verified to be isolated on a periodic basis. This is necessary to ensure that containment penetrations required to be isolated following an accident and no longer capable of being automatically isolated will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification, through a system walkdown or computer status indication, that those isolation devices outside containment and capable of being mispositioned are in the correct position. The Completion Time of "once per 31 days for isolation devices outside containment" is appropriate considering the fact that the devices are operated under administrative controls and the probability of their

misalignment is low. For the isolation devices inside containment, the time period specified as "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is* based on engineering judgment and is considered reasonable in view of the inaccessibility of the isolation devices and other administrative controls that will ensure that isolation device misalignment is an unlikely possibility.

Condition A has been modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two containment Catawba Units 1 and 2 B 3.6.3-6 Revision No. 5

Containment Isolation Valves B 3.6.3

BASES ACTIONS (continued) isolation valves. For penetration flow paths with only one containment isolation valve and a closed system, Condition C provides the appropriate actions.

Required Action A.2 is modified by a Note. that applies to isolation devices located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these devices once they have been verified to be in the proper position, is small.

With two containment isolation valves in one or more penetration flow paths inoperable, the affected penetration flow path must be isolated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-

activated automatic valve, a closed manual valve, and a blind flange .

The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time is consistent with the ACTIONS of LCO 3.6.1. In the event the affected penetration is isolated in accordance with Required Action B.1, the affected penetration must be verifie~ to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verifiGation is necessary to assure leak tightness of containment and that penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration flow path is isolated is appropriate considering the fact that the valves are operated under administrative control and the probability of their misalignment is low.

Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two containment isolation valves.

Condition A of this L,CO addresses the condition of one containment isolation valve inoperable in this type of penetration flow path .

Catawba Units 1 and 2 B 3.6.3-7 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued)

C.1 and C.2 With one or more penetration flow paths with one containment isolation valve inoperable, the inoperable valve flow path must be restored to OPERABLE status or the affected penetration flow path must be isolated.

The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure.

Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration flow path.

Required Action C.1 must be completed within the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time. The specified time period is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of maintaining containment integrity during MODES 1, 2, 3, and 4. In the event the affected penetration flow path is isolated in accordance with Required Action C.1, the affected penetration flow path must be verified to be isolated on a periodic basis. This periodic verification is necessary to assure leak tightness of containment and that containment penetrations requiring isolation following an accident are isolated. The Completion Time of once per 31 days for verifying that each affected penetration flow path is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.

Condition C is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with only one containment isolation valve and a closed system. The closed system must meet the requirements of Reference 4. This Note is necessary since this Condition is written to specifically address those penetration flow paths in a closed system.

Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have been V,erified to be in the proper position, is small.

Catawba Units 1 and 2 B 3.6.3-8 Revision No. 5

Containment Isolation Valves B 3.6.3

BASES ACTIONS (continued)

With the reactor building bypass leakage rate not within limit, the assumptions of the safety analyses are not met. Therefore, the leakage must be restored to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Restoration can be accomplished by isolating the penetration(s) that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. When a penetration is isolated the leakage rate for the isolated penetration is assumed .to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actu.al pathway leakage of the two devices. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration(s) and the relative importance of secondary containment bypass leakage to the overall containment function.

E.1, E.2. and E.3 In the event one or more containment purge, hydrogen purge, or

containment air release and addition valves in one or more penetration

. flow paths are not within the leakage limits, leakage must be restored to within limits, or the affected penetration flow path must be isolated. The method of isolation must be by the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, closed manual valve, or blind flange. A valve with resilient seals utilized to satisfy Required Action E.1 must have been demonstrated to meet the leakage requirements of SR 3.6.3.6. The specified Completion

. Time is reasonable, considering that one. containment purge valve remains closed so that a gross breach of containment does not exist.

In accordance with Required Action E.2, this penetration flow path must be verified to be isolated on a periodic basis. The periodic verification is necessary to ensure that containment penetrations required to be isolated following an accident, which are no longer capable of being automatically isolated, will be in the isolation position should an event occur. This Required Action does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown or computer status indication, that those isolation d.evices outside containment capable of being mispositioned are in the correct position .

Catawba Units 1 and 2 B 3.6.3-9 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued)

For the isolation devices inside containment, the time period specified as "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the isolation devices and other administrative controls that will ensure that isolation device misalignment is an unlikely possibility.

For the valve with resilient seal that is isolated in accordance with Required Action E.1, SR 3.6.3.6 must be performed at least once every 92 days. This assures that degradation of the resilient seal is detected and confirms that the leakage rate of the containment purge valve does not increase during the time the penetration is isolated.

F.1 and F.2 If the Required Actions and associated Completion Times are not met, the plant must be brought to a MODE in which the LCO does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to 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 plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.3.1 REQUIREMENTS Each containment purge supply and exhaust isolation valve for the lower compartment and the upper compartment, instrument room, and the Hydrogen Purge System is required to be verified sealed closed. This Surveillance is designed to ensure that a gross breach of containment is not caused by an inadvertent or spurious opening of a containment purge valve. Detailed analysis of these valves to conclusively demonstrate their ability to close during a LOCA in time to limit offsite doses has not been performed. Therefore, these valves are required to be in the sealed closed position during MODES 1, 2, 3, and 4. A valve that is sealed closed must have motive power to the valve operator removed. This can be accomplished by de-energizing the source of electric power or by removing the air supply to the valve operator. In this application, the term "sealed" has no connotation of leak tightness.

Catawba Units 1 and 2 B 3.6.3-10 Revision No. 5

Containment Isolation Valves B 3.6.3

BASES SURVEILLANCE REQUIREMENTS (continued)

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. In the event valve leakage requires entry into Condition E, the Surveillance permits opening one valve in a penetration flow path to perform repairs.

SR 3.6.3.2 This SR ensures that the Containment Air Release and Addition System isolation valves are closed as required or, if open, open for an allowable reason. If a valve is open in violation of this SR, the valve is considered inoperable. If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of limits. The SR is not required to be met when the valves are open for the reasons stated. The val.ves may be opened for pressure control, ALARA or air quality considerations for personnel entry, or for

. Surveillances that require the valves to* be open. The valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The Surveillance

. Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident

. leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through system walkdown or computer status indication, that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open .

Catawba Units 1 and 2 B 3.6.3-11 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)

This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be the correct position upon locking, sealing, or securing.

The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3 and 4 for AU\RA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.

SR 3.6.3.4 This SR requires verification that each containment isolation manual valve and blind flange located inside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked; sealed, or otherwise secured in the closed position, since these were verified to be the correct position upon locking, sealing, or securing.

This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification*

by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment

isolation valves, once they have been verified to be in their proper position, is small.

Catawba Units 1 and 2 B 3.6.3-12 Revision No. 5

Containment Isolation Valves B 3.6.3

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.3.5 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.

The isolation time is specified in the UFSAR and the Frequency of this SR is in accordance with the lnservice Testing Program .

. SR 3.6.3.6 For the Containment Purge System valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B is required to ensure OPERABILITY. The measured leakage rate for the Containment Purge System and Hydrogen Purge System valves must be~ 0.05 La when pressurized to Pa.

Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than other seal types. Based on this observation and the importance of maintaining this penetration

leak tight (due to the direct path between containment and the environment), these valves will not be placed on the maximum extended test interval. Therefore, these valves will be tested in accordance with NEI 94-01 with a maximum test interval of 30 months.

The Containment Air Release and Addition System valves have a demonstrated history of acceptable leakage. The measured leakage rate for containment air release and addition valves must be~ 0.01 La when pressurized to Pa. These valves will be tested in accordance with NEI 94-01 with a maximum test interval of 30 months.

SR 3.6.3.7 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a OBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment

Catawba Units 1 and 2 B 3.6.3-13 Revision No. 5

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued) isolation signal. The isolation signals involved are Phase A, Phase B, and Safety Injection. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6.3.8 This SR ensures that the combined leakage rate of all reactor building bypass leakage paths is less than or equal to the specified leakage rate.

This provides assurance that the assumptions in the safety analysis are met. The Frequency is required by the Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria.

Bypass leakage is considered part of La.

REFERENCES 1. UFSAR, Section 15.

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

3. UFSAR, Section 6.2.

4. Standard Review Plan 6.2.4.

5. Not used.

Catawba Units 1 and 2 B 3.6.3-14 Revision No. 5

Containment Spray System B 3.6.6

B 3.6 CONTAINMENT SYSTEMS B 3~6.6 Containment Spray System BASES --z BACKGROUND The Containment Spray System provides containment atmosphere cooling to limit post accident pressure and temperature in containment to less than the design values. Reduction of containment pressure and the iodine removal capability of the spray reduce the release of fission product radioactivity from containment to the environment, in the event of a Design Basis Accident (DBA). The Containment Spray System is designed to meet the requirements of 10 CFR 50, Appendix A, GDC 38, "Containment Heat Removal," GDC 39, "Inspection of Containment Heat Removal Systems," GDC 40, "Testing of Containment Heat Removal Systems," GDC 41, "Containment Atmosphere Cleanup," GDC 42,

"Inspection of Containment Atmosphere Cleanup Systems," and GDC 43, "Testing of Containment Atmosphere Cleanup Systems" (Ref. 1).

The Containment Spray System consists of two separate trains of equal capacity, each capable of meeting the system design basis spray

coverage. Each train includes a containment spray pump, one containment spray heat exchanger, spray headers, nozzles, valves, and piping. Each train is powered from a separate Engineered Safety Feature (ESF) bus. The containment spray pumps are started manually with the pump suction aligned to the containment sump once the ECCS is aligned to the recirculation mode of operation.

The diversion of a portion of the recirculation flow from each train of the--

Residual Heat Removal (RHR) System to additional redundant spray heacterscompletes the Containment Spray System heat removal capability. Each RHR train is capable of supplying spray coverage, if desired, to supplement the Containment Spray System.

The Containment Spray System provides a spray of cold or subcooled borated water into the upper containment volume to limit the containment pressure and temperature during a DBA. In the recirculation mode of operation, heat is removed from the containment sump water by the Containment Spray System and RHR heat exchangers. Each train of the Containment Spray System, provides adequate spray coverage to meet the system design requirements for containment heat removal.

Catawba Units 1 and 2 B 3.6.6-1 Revision No. 7

Containment Spray System B 3.6.6 BASES BACKGROUND (continued)

For the hypothetical double-ended rupture of a Reactor Coolant System pipe, the pH of the sump solution (and, consequently, the spray solution) is raised to approximately 7.9 within one hour of the onset of the LOCA.

The resultant pH of the sump solution is based on the mixing of the RCS fluids, ECCS injection fluid, and the melted ice which are combined in the sump. The alkaline pH of the containment sump water minimizes the evolution of iodine and the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to the fluid.

The Containment Spray System is actuated manually. The Containment Spray System is not actuated until an RWST level Low-Low alarm is received and the ECCS has been realigned in the recirculation mode of operation. The Low-Low alarm for the RWST signals the operator to manually align the ECCS to the recirculation mode and to manually initiate the Containment Spray System. The Containment Spray System maintains an equilibrium temperature between the containment atmosphere and the recirculated sump water. Operation of the Containment Spray System in the recirculation mode is controlled by the operator in accordance with the emergency operation procedures.

The RHR spray operation is initiated manually, when required by the

emergency operating procedures, after the ECCS is operating in the recirculation mode. The RHR sprays are available to supplement the Containment Spray System, if desired, in limiting containment pressure.

This additional spray capacity would typically be used after the ice bed has been depleted and in the event that containment pressure rises above a predetermined limit. However, RHR spray operation is not credited for design basis events. The Containment Spray System is designed to ensure that the heat removal capability required during the post accident period can be attained.

The operation of the Containment Spray System, together with the ice condenser, is adequate to assure pressure suppression subsequent to the initial blowdown of steam and water from a OBA. During the post blowdown period, the Air Return System (ARS) is automatically started.

The ARS returns upper compartment air through the divider barrier to the lower compartment. This serves to equalize pressures in containment and to continue circulating heated air and steam through the ice condenser, where heat is removed by the remaining ice.

The Containment Spray System limits the temperature and pressure that could be expected following a OBA. Protection of containment integrity limits leakage of fission product radioactivity from containment to the environment.

Catawba Units 1 and 2 B 3.6.6-2 Revision No. 7

Containment Spray System B 3.6.6

BASES APPLICABLE The limiting OBAs considered relative to containment OPERABILITY SAFETY ANALYSES are the loss of coolant accident (LOCA) and the steam line break (SLB).

The OBA LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients.

No two OBAs are assumed to occur simultaneously or consecutively.

The postulated OBAs are analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train *Of the Containment Spray System, the RHR System, and the ARS being rendered inoperable (Ref. 2).

The OBA analyses show that the maximum peak containment pressure results from the LOCA analysis, and is calculated to be less than the containment design pressure. The maximum peak containment atmosphere temperature results from the SLB analysis and was calculated to be within the containment environmental qualification temperature during the OBA SLB .. The basis of the containment environmental qualification temperature is to ensure the OPERABILITY of safety related equipment inside containment (Ref. 3).

The Containment Spray System actuation modeled in the containment analysis is based on the time associated with reaching the RWST low

level setpoint prior to achieving full flow through the containment spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The Containment Spray System total response time is composed of operator action delay and system startup time.

For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the ECCS cooling effectiveness during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with 10 CFR 50, Appendix K (Ref. 4).

Inadvertent actuation is precluded by a design feature consisting of an additional set of containment pressure sensors which prevents operation when the containment pressure is below the containment pressure control system permissive.

The Containment Spray System satisfies Criterion 3 of 10 CFR 50.36 (Ref. 5) .

Catawba Units 1 and 2 B 3.6.6-3 Revision No. 7

Containment Spray System B 3.6.6 BASES LCO During a OBA, one train of Containment Spray System is required to provide the heat removal capability assumed in the safety analyses. To ensure that this requirement is met, two containment spray trains must be OPERABLE with power from two safety related, independent power supplies. Therefore, in the event of an accident, at least one train operates.

Each Containment Spray System includes a spray pump, headers, valves, heat exchangers, nozzles, piping, instruments, and controls to ensure an OPERABLE flow path capable of being manually initiated to take suction from the containment sump and delivering it to the containment spray rings. Management of gas voids is important to Containment Spray System OPERABILITY.

APPL! GABI LITY In MODES 1, 2, 3, and 4; a OBA could cause a release of radioactive material to containment and an increase in containment pressure and temperature requiring the operation of the Containment Spray System.

In MODES 5 and 6, the probability and consequences of these events are reduced because of the pressure and temperature limitations of these MODES. Thus, the Containment Spray System is not required to be OPERABLE in MODE 5 or 6.

ACTIONS With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The components in this degraded condition are capable of providing 100% of the heat removal after an accident. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time was developed taking into account the redundant heat removal and iodine removal capabilities afforded by the OPERABLE train and the low probability of a OBA occurring during this period.

Catawba Units 1 and 2 B 3.6.6-4 Revision No. 7

Containment Spray System B 3.6.6

BASES ACTIONS (continued)

B.1 and B.2 If the affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.

SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Ve,rifying the correct alignment of manual and power operated valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System

operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown or computer status indication, that those valves outside containment and capable of potentially being mispositioned, are in the correct position. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

The Surveillance is modified by a Note which exempts system vent flow paths opened under administrative control. *The administrative control should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication with the operators in the control room. This individual will have a method to rapidly close the system vent flow path if directed.

SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded during the cycle.

Flow and differential head are normal tests of centrifugal pump

Catawba Units 1 and 2 B 3.6.6-5 Revision No. 7

Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) performance required by the ASME Code (Ref. 6). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the lnservice Testing Program.

SR 3.6.6.3 and SR 3.6.6.4 Not used.

SR 3.6.6.5 and SR 3.6.6.6 These SRs require verification of proper interaction between the CPCS system and the Containment Spray System.

SR 3.6.6.5 deals solely with the containment spray pumps. It must be

shown through testing that: (1) the containment spray pumps are prevented from starting in the absence of a CPCS permissive, (2) the containment spray pumps can be manually started when given a CPCS permissive, and (3) when running, the containment spray pumps stop when the CPCS permissive is removed. The "inhibit", "permit", and "terminate" parts of the CPCS interface with the containment spray pumps are verified by testing in this fashion.

SR 3.6.6.6 deals solely with containment spray header containment isolation valves NS12B, NS158, NS29A, and NS32A. It must be shown through testing that: (1) each valve closes when the CPCS permissive is removed, OR (2) each valve is prevented from opening in the absence of a CPCS permissive. In addition to one of the above, it must also be shown that each valve can be manually opened when given a CPCS permissive.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Catawba Units 1 and 2 B 3.6.6-6 Revision No. 7

Containment Spray System B 3.6.6

BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.6.7 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. The spray nozzles can also be tested using a vacuum blower to induce air flow through each nozzle to verify unobstructed flow.

This SR requires verification that each spray nozzle is unobstructed following activities that could cause nozzle blockage. Normal plant operation and activities are not expected to initiate this SR. However, activities such as inadvertent spray actuation that causes fluid flow through the nozzles, major configuration change, or a loss of foreign material control when working within the respective system boundary may require Surveillance performance.

SR 3.6.6.8 Containment Spray System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the

containment spray trains and may also prevent water hammer and pump cavitation.

Selection of Containment Spray System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration.

Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The Containment Spray System is OPERABLE when it is sufficiently filled with water. Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump), the Surveillance is not met. If it is determined by subsequent evaluation that the Containment Spray System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met.

Catawba Units 1 and 2 B 3.6.6-7 Revision No. 7

Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued)

Accumulated gas should be eliminated or brought within the acceptance criteria limits.

Containment Spray System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations. Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to mqnitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the

Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 38, GDC 39, GDC 40, GDC 41, GDC 42, and GDC 43.

2. UFSAR, Section 6.2.

3. 10 CFR 50.49.

4. 10 CFR 50, Appendix K.

5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

6. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Catawba Units 1 and 2 B 3.6.6-8 Revision No. 7

Nuclear Instrumentation B 3.9.2

B 3.9 REFUELING OPERATIONS B 3.9.2 Nuclear Instrumentation BASES BACKGROUND The neutron flux monitors are used during refueling operations to monitor the core reactivity condition. The installed neutron flux monitors are part of the Nuclear Instrumentation System (NIS) and the Wide Range Neutron Flux Monitoring System (Gamma-Metrics). Source range indication is provided via the N IS source range channels and the Gamma-Metrics shutdown monitors using detectors located external to the reactor vessel.

These detectors monitor neutrons leaking from the core. Neutron flux indication is provided in counts per second. The NIS source range channels have a range of 0.1 to 1E6 cps. The wide range channels have a range of 0.1 to 1E5 cps (in the startup range). The NIS source range channels and the Gamma-Metrics shutdown monitors provide continuous visible count rate indication in the control room. The NIS is designed in accordance with the criteria presented in Reference 1.

The shutdown monitors (Gamma-Metrics) automatic actuations and alarm are not required for OPERABILITY during refueling operations. The NIS source range audible indication and audible alarm are not required for OPERABILITY during refueling operations.

APPLICABLE Two OPERABLE neutron flux monitors are required to provide SAFETY an indication to alert the operator to unexpected changes in core ANALYSES reactivity such as with a boron dilution accident (Ref. 2) or an improperly loaded fuel assembly.

The neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36 (Ref. 3).

LCO This LCO requires that two neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE, each monitor must provide visual indication .

Catawba Units 1 and 2 B 3.9.2-1 Revision No. 6

Nuclear Instrumentation B 3.9.2 BASES APPLICABILITY In MODE 6, the neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODES 2, 3, 4, and 5, the NIS source range detectors and circuitry are also required to be OPERABLE by LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." The Gamma-Metrics shutdown monitors are required to be OPERABLE in MODES 3, 4 and 5 by LCO 3.3.9, "Boron Dilution Mitigation System (BDMS)."

ACTIONS A.1 and A.2 With only one required neutron flux monitor OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.1 must be suspended immediately. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation. Introduction of coolant inventory must be from sources that have a boron concentration greater than that which would be required in the RCS for minimum refueling boron concentration. This may result in overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. Performance of Required Action A.1 shall not preclude

completion of movement of a component to a safe position.

With no required neutron flux monitor OPERABLE, action to restore a monitor to OPERABLE status shall be initiated immediately. Once initiated, action shall be continued until a neutron flux monitor is restored to OPERABLE status.

With no neutron flux monitor OPERABLE, there are no direct means of detecting ch_anges in core reactivity. However, since CORE ALTERATIONS and introduction of coolant into the RCS with boron concentration less than required to meet the minimum boron concentration of LCO 3.9.1 must be suspended immediately, the core reactivity condition is stabilized until the neutron flux monitors are OPERABLE. This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.

Catawba Units 1 and 2 B 3.9.2-2 Revision No. 6

Nuclear Instrumentation B 3.9.2 BASES ACTIONS (continued)

The Completion Time of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to obtain and analyze-a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.

SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program .

SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION. This SR is modified by a Note stating that neutron detector sensors (NIS and BDMS) are excluded from the CHANNEL CALIBRATION.

The CHANNEL CALIBRATION for the fission chamber source range neutron detectors and for the source range neutron flux monitors (Gamma-Metrics) consists of verifying that the channels respond correctly to test inputs with the necessary range and accuracy.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 13, GDC 26, GDC 28, and GDC 29.

2. UFSAR, Sections 4.2, 15.4.6.

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

Catawba Units 1 and 2 B 3.9.2-3 Revision No. 6

RHR and Coolant Circulation-High Water Level B 3.9.4

B 3.9 REFUELING OPERATIONS B 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level BASES BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant and to prevent boron stratification (Ref. 1). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchanger(s), where the heat is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR System for normal cooldown or decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant and component cooling water through the RHR heat exchanger(s). Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR System.

APPLICABLE If the reactor coolant temperature is not maintained below 200°F, boiling SAFETY ANALYSES of the reactor coolant could result. This could lead to a loss of coolant in the reactor vessel. Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. One train of the RHR System is required to be operational in MODE 6, with the water level ~ 23 ft above the top of the reactor vessel flange, to prevent this challenge. The LCO does permit de-energizing the RHR pump for short durations, under the condition that the boron concentration is not diluted. This conditional de-energizing of the RHR pump does not result in a challenge to the fission product barrier.

The RHR System satisfies Criterion 4 of 10 CFR 50.36 (Ref. 2).

LCO Only one RHR loop is required for decay heat removal in MODE 6, with the water level ~ 23 ft above the top of the reactor vessel flange. Only one RHR loop is required to be OPERABLE, because the volume of water above the reactor vessel flange provides backup decay heat removal capability. At least one RHR loop must be OPERABLE and in operation to provide:

Catawba Units 1 and 2 B 3.9.4-1 Revision No. 5

RHR and Coolant Circulation-High Water Level B 3.9.4 BASES LCO (continued)

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility of criticality; and

c. Indication of reactor coolant temperature.

An OPERABLE RHR loop includes an RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The operability of the operating RHR train and the supporting heat sink is dependent on the ability to maintain the desired RCS temperature. If not in its normal RHR alignment from the RCS hot leg and returning to the RCS cold legs, the required RHR loop is OPERABLE provided the system may be placed in service from the control room, or may be placed in service in a short period of time by actions outside the control room and there are no restraints to placing the equipment in service. Management of gas voids is important to RHR System OPERABILITY.

The LCO is modified by a Note that allows the required operating RHR loop to be removed from service for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period,

provided no operations are permitted that would dilute the RCS boron concentration with coolant at boron concentrations less than required to meet the minimum boron concentration of LCO 3.9.1. Boron concentration reduction with coolant at boron concentrations less than required to assure the minimum required RCS boron concentration is maintained is prohibited because uniform concentration distribution cannot be ensured without forced .circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve testing. During this 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.

The acceptability of the LCO and the LCO Note is based on preventing boiling in the core in the event of the loss of RHR cooling. However, it has been determined that when the upper internals are in place in the reactor vessel there is insufficient communication with the water above the core for adequate decay heat removal by natural circulation. As a result, boiling in the core could occur in a relatively short time if RHR cooling is lost. Therefore, during the short period of time that the upper internals are installed, administrative processes are implemented to reduce the risk of core boiling. The availability of additional cooling equipment, including equipment not required to be OPERABLE by the Technical Specifications, contributes to this risk reduction. The plant staff assesses these cooling sources to assure that the desired minimal level of risk is maintained .

Catawba Units 1 and 2 B 3.9.4-2 Revision No. 5

RHR and Coolant Circulation-High Water Level B 3.9.4

BASES LCO (continued)

This is commonly referred to as defense-in-depth. This strategy is consistent with NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management" (Ref. 3).

APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6, with the water level ~ 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movement in LCO 3.9.6, "Refueling Cavity Water Level." Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS), and- Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level

< 23 ft are located in LCO 3.9.5, '*'Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level."

ACTIONS RHR loop requirements are met by having one RHR loop OPERABLE and in operation, except as permitted in the Note to the LCO .

If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation.

Introduction of coolant inventory must be from sources that have a boron concentration greater th.an that which would be required in the RCS for

minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

If RHR loop requirements are not met, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core.

A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition .

Catawba Units 1 and 2 B 3.9.4-3 Revision No. 5

RHR and Coolant Circulation-High Water Level B 3.9.4 BASES ACTIONS (continued)

If RHR loop requirements are not met, actions shall be initiated and continued in order to satisfy RHR loop requirements. With the unit in MODE 6 and the refueling water level ;:::: 23 ft above the top of the reactor vessel flange, corrective actions shall be initiated immediately.

If RHR loop requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on the low probability of the coolant boiling in that time.

SURVEILLANCE REQUIREMENTS SR 3.9.4.1 This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The RCS temperature is determined to ensure the appropriate decay heat removal is maintained. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.9.4.2 RHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR loops and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of RHR System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation drawings, and calculations. The design review is supplemented by Catawba Units 1 and 2 B 3.9.4-4 Revision No. 5

RHR and Coolant Circulation-High Water Level B 3.9.4

BASES SURVEILLANCE REQUIREMENTS (continued) system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump),

the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance

criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation .

Catawba Units 1 and 2 B 3.9.4-5 Revision No. 5

RHR and Coolant Circulation-High Water Level B 3.9.4 BASES REFERENCES 1.

2.

UFSAR, Section 5.5.7.

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

3. NUMARC 91-06, "Guidelines for Industry Actions to Assess Shutdown Management."

Catawba Units 1 and 2 B 3.9.4-6 Revision No. 5

RHR and Coolant Circulation-Low Water Level B 3.9.5

B 3.9 REFUELING OPERATIONS B 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System (RCS), as required by GDC 34, to provide mixing of borated coolant, and to prevent boron stratification (Ref. 1). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg(s). Operation of the RHR System for normal cooldown decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant and component cooling water through the RHR heat exchanger(s). Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the RHR System.

APPLICABLE If the reactor coolant temperature is not maintained below 200°F, boiling SAFETY ANALYSES of the reactor coolant could result This could lead to a loss of coolant in the reactor vessel. Additionally, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant will eventually challenge the integrity of the fuel cladding, which is a fission product barrier. Two trains of the RHR System are required to be OPERABLE, and one train in operation, in order to prevent this challenge.

The RHR System satisfies Criterion 4of10 CFR 50.36 (Ref. 2).

LCO In MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, both RHR loops must be OPERABLE.

Additionally, one loop of RHR must be in operation in order to provide:

a. Removal of decay heat;

b. Mixing of borated coolant to minimize the possibility of criticality; and

c. Indication of reactor coolant temperature .

Catawba Units 1 and 2 B 3.9.5-1 Revision No. 4

RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES LCO (continued)

An OPERABLE RHR loop consists of an RHR pump, a heat exchanger, valves, piping, instruments and controls to ensure an OPERABLE flow path and to determine the low end temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. The operability of the operating RHR train and the supporting heat sink is dependent on the ability to maintain the desired RCS temperature. If not in its normal RHR alignment from the RCS hot leg and returning to the RCS cold legs, the required RHR loop is OPERABLE provided the system may be placed in service from the control room, or may be placed in service in a short period of time by actions outside the control room and there are no restraints to placing the equipment in service. Management of gas voids is important to RHR System OPERABILITY.

Both RHR pumps may be aligned to the Refueling Water Storage Tank to support filling the refueling cavity or for performance of required testing.

APPLICABILITY Two RHR loops are required to be OPERABLE, and one RHR loop must be in operation in MODE 6, with the water level < 23 ft above the top of the reactor vessel flange, to provide decay heat removal. Requirements for the RHR System in other MODES are covered by LCOs in Section 3.4, Reactor Coolant System (RCS}, and Section 3.5, Emergency Core Cooling Systems (ECCS). RHR loop requirements in MODE 6 with the water level ;:::: 23 ft are located in LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation-High Water Level."

ACTIONS A.1 and A.2 If less than the required number of RHR loops are OPERABLE, action shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until ;:::: 23 ft of water level is established above the reactor vessel flange. When the water level is

23 ft above the reactor vessel flange, the Applicability changes to that of LCO 3.9.4, and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.

If no RHR loop is in operation, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron Catawba Units 1 and 2 B 3.9.5-2 Revision No. 4

RHR and Coolant Circulation-Low Water Level B 3.9.5

BASES ACTIONS (continued) concentration limit is required to assure continued safe operation.

Introduction of coolant inventory must be from sources that have a boron concentration greater than that which would be required in the RCS for minimum refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation.

If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation. Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.

If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR loop

requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is appropriate for the majority of time during refueling operations, based on time to coolant boiling, since water level is not routinely maintained at low levels.

SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability, prevent vortexing in the suction of the RHR pumps, and to prevent thermal and boron stratification in the core. The RCS temperature is determined to ensure the appropriate decay heat removal is maintained. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program .

Catawba Units 1 and 2 B 3.9.5-3 Revision No. 4

RHR and Coolant Circulation-Low Water Level B 3.9.5 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.9.5.3 RHR System piping and components have the potential to develop voids and pockets of entrained gases. Preventing and managing gas intrusion and accumulation is necessary for proper operation of the RHR loops and may also prevent water hammer, pump cavitation, and pumping of noncondensible gas into the reactor vessel.

Selection of RHR System locations susceptible to gas accumulation is based on a review of system design information, including piping and instrumentation drawings, isometric drawings, plan and elevation

drawings, and calculations. The design review is supplemented by system walk downs to validate the system high points and to confirm the location and orientation of important components that can become sources of gas or could otherwise cause gas to be trapped or difficult to remove during system maintenance or restoration. Susceptible locations depend on plant and system configuration, such as stand-by versus operating conditions.

The RHR System is OPERABLE when it is sufficiently filled with water.

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. If accumulated gas is discovered that exceeds the acceptance criteria for the susceptible location (or the volume of accumulated gas at one or more susceptible locations exceeds an acceptance criteria for gas volume at the suction or discharge of a pump),

the Surveillance is not met. If it is determined by subsequent evaluation that the RHR System is not rendered inoperable by the accumulated gas (i.e., the system is sufficiently filled with water), the Surveillance may be declared met. Accumulated gas should be eliminated or brought within the acceptance criteria limits.

RHR System locations susceptible to gas accumulation are monitored and, if gas is found, the gas volume is compared to the acceptance criteria for the location. Susceptible locations in the same system flow path which are subject to the same gas intrusion mechanisms may be Catawba Units 1 and 2 B 3.9.5-4 Revision No. 4

RHR and Coolant Circulation-Low Water Level B 3.9.5

BASES SURVEILLANCE REQUIREMENTS (continued) verified by monitoring a representative sub-set of susceptible locations.

Monitoring may not be practical for locations that are inaccessible due to radiological or environmental conditions, the plant configuration, or personnel safety. For these locations alternative methods (e.g., operating parameters, remote monitoring) may be used to monitor the susceptible location. Monitoring is not required for susceptible locations where the maximum potential accumulated gas void volume has been evaluated and determined to not challenge system OPERABILITY. The accuracy of the method used for monitoring the susceptible locations and trending of the results should be sufficient to assure system OPERABILITY during the Surveillance interval.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. The Surveillance Frequency may vary by location susceptible to gas accumulation.

REFERENCES 1. UFSAR, Section 5.5.7.

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

Catawba Units 1 and 2 B 3.9.5-5 Revision No. 4

CNS-17-033, Technical Specification Bases Changes

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