ML22067A030

Changes to Technical Specification Bases - Revisions 86, 87, 88, 89, and 90

ML22067A030
Person / Time
Site:	Wolf Creek
Issue date:	03/08/2022
From:	Benham R Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 22-0023
Download: ML22067A030 (266)

Text

Ron Benham Director Nuclear and Regulatory Affairs March 08, 2022 RA 22-0023 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Docket No. 50-482: Wolf Creek Generating Station Changes to Technical Specification Bases - Revisions 86, 87, 88, 89, and 90 Commissioners and Staff:

The Wolf Creek Generating Station (WCGS) Unit 1 Technical Specifications (TS), Section 5.5.14, Technical Specifications (TS) Bases Control Program, provides the means for making changes to the Bases without prior Nuclear Regulatory Commission (NRC) approval. In addition, TS Section 5.5.14 requires that changes made without NRC approval be provided to the NRC on a frequency consistent with 10 CFR 50.71(e). The Enclosure provides those changes made to the WCGS TS Bases (Revisions 86, 87, 88, 89, and 90) under the provisions to TS Section 5.5.14 and a List of Effective Pages. This submittal reflects changes from January 1, 2021, through December 31, 2021.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4204.

Sincerely, Ron Benham RDB/rlt

Enclosure:

Wolf Creek Generating Station Changes to the Technical Specification Bases cc: S. S. Lee (NRC), w/e S. A. Morris (NRC), w/e G. E. Werner (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 l Burlington, KS 66839 l 620-364-8831

Enclosure to RA 22-0023 Wolf Creek Generating Station Changes to the Technical Specification Bases (264 pages)

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TAB - Title Page Technical Specification Cover Page Title Page TAB - Table of Contents i 34 DRR 07-1057 7/10/07 ii 29 DRR 06-1984 10/17/06 iii 44 DRR 09-1744 10/28/09 TAB - B 2.0 SAFETY LIMITS (SLs)

B 2.1.1-1 0 Amend. No. 123 12/18/99 B 2.1.1-2 14 DRR 03-0102 2/12/03 B 2.1.1-3 14 DRR 03-0102 2/12/03 B 2.1.1-4 0 Amend. No. 123 2/12/03 B 2.1.2-1 84 DRR 20-0400 08/18/20 B 2.1.2-2 84 DRR 20-0400 08/18/20 B 2.1.2-3 81 DRR 19-1027 10/28/19 TAB - B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 81 DRR 19-1027 10/28/19 B 3.0-2 0 Amend. No. 123 12/18/99 B 3.0-3 81 DRR 19-1027 10/28/19 B 3.0-4 81 DRR 19-1027 10/28/19 B 3.0-5 81 DRR 19-1027 10/28/19 B 3.0-6 81 DRR 19-1027 10/28/19 B 3.0-7 81 DRR 19-1027 10/28/19 B 3.0-8 81 DRR 19-1027 10/28/19 B 3.0-9 81 DRR 19-1027 10/28/19 B 3.0-10 81 DRR 19-1027 10/28/19 B 3.0-11 81 DRR 19-1027 10/28/19 B 3.0-12 81 DRR 19-1027 10/28/19 B 3.0-13 81 DRR 19-1027 10/28/19 B 3.0-14 81 DRR 19-1027 10/28/19 B 3.0-15 81 DRR 19-1027 10/28/19 B 3.0-16 81 DRR 19-1027 10/28/19 B 3.0-17 81 DRR 19-1027 10/28/19 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1-1 0 Amend. No. 123 12/18/99 B 3.1.1-2 0 Amend. No. 123 12/18/99 B 3.1.1-3 0 Amend. No. 123 12/18/99 B 3.1.1-4 81 DRR 19-1027 10/28/19 B 3.1.1-5 81 DRR 19-1027 10/28/19 B 3.1.2-1 0 Amend. No. 123 12/18/99 B 3.1.2-2 0 Amend. No. 123 12/18/99 B 3.1.2-3 0 Amend. No. 123 12/18/99 B 3.1.2-4 0 Amend. No. 123 12/18/99 B 3.1.2-5 0 Amend. No. 123 12/18/99 B 3.1.3-1 0 Amend. No. 123 12/18/99 B 3.1.3-2 0 Amend. No. 123 12/18/99 B 3.1.3-3 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 i Revision 86

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TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)

B 3.1.3-4 0 Amend. No. 123 12/18/99 B 3.1.3-5 0 Amend. No. 123 12/18/99 B 3.1.3-6 0 Amend. No. 123 12/18/99 B 3.1.4-1 0 Amend. No. 123 12/18/99 B 3.1.4-2 0 Amend. No. 123 12/18/99 B 3.1.4-3 48 DRR 10-3740 12/28/10 B 3.1.4-4 0 Amend. No. 123 12/18/99 B 3.1.4-5 0 Amend. No. 123 12/18/99 B 3.1.4-6 48 DRR 10-3740 12/28/10 B 3.1.4-7 0 Amend. No. 123 12/18/99 B 3.1.4-8 0 Amend. No. 123 12/18/99 B 3.1.4-9 0 Amend. No. 123 12/18/99 B 3.1.5-1 0 Amend. No. 123 12/18/99 B 3.1.5-2 0 Amend. No. 123 12/18/99 B 3.1.5-3 0 Amend. No. 123 12/18/99 B 3.1.5-4 0 Amend. No. 123 12/18/99 B 3.1.6-1 0 Amend. No. 123 12/18/99 B 3.1.6-2 0 Amend. No. 123 12/18/99 B 3.1.6-3 0 Amend. No. 123 12/18/99 B 3.1.6-4 0 Amend. No. 123 12/18/99 B 3.1.6-5 0 Amend. No. 123 12/18/99 B 3.1.6-6 0 Amend. No. 123 12/18/99 B 3.1.7-1 0 Amend. No. 123 12/18/99 B 3.1.7-2 0 Amend. No. 123 12/18/99 B 3.1.7-3 48 DRR 10-3740 12/28/10 B 3.1.7-4 48 DRR 10-3740 12/28/10 B 3.1.7-5 48 DRR 10-3740 12/28/10 B 3.1.7-6 0 Amend. No. 123 12/18/99 B 3.1.8-1 0 Amend. No. 123 12/18/99 B 3.1.8-2 0 Amend. No. 123 12/18/99 B 3.1.8-3 15 DRR 03-0860 7/10/03 B 3.1.8-4 15 DRR 03-0860 7/10/03 B 3.1.8-5 0 Amend. No. 123 12/18/99 B 3.1.8-6 5 DRR 00-1427 10/12/00 B 3.1.9-1 84 DRR 20-0400 08/18/20 B 3.1.9-2 84 DRR 20-0400 08/18/20 B 3.1.9-3 84 DRR 20-0400 08/18/20 B 3.1.9-4 84 DRR 20-0400 08/18/20 B 3.1.9-5 84 DRR 20-0400 08/18/20 TAB - B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1-1 48 DRR 10-3740 12/28/10 B 3.2.1-2 0 Amend. No. 123 12/18/99 B 3.2.1-3 48 DRR 10-3740 12/28/10 B 3.2.1-4 48 DRR 10-3740 12/28/10 B 3.2.1-5 48 DRR 10-3740 12/28/10 B 3.2.1-6 48 DRR 10-3740 12/28/10 B 3.2.1-7 48 DRR 10-3740 12/28/10 B 3.2.1-8 48 DRR 10-3740 12/28/10 B 3.2.1-9 29 DRR 06-1984 10/17/06 B 3.2.1-10 70 DRR 15-0944 4/28/15 Wolf Creek - Unit 1 ii Revision 86

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TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued)

B 3.2.2-1 48 DRR 10-3740 12/28/10 B 3.2.2-2 0 Amend. No. 123 12/18/99 B 3.2.2-3 48 DRR 10-3740 12/28/10 B 3.2.2-4 48 DRR 10-3740 12/28/10 B 3.2.2-5 48 DRR 10-3740 12/28/10 B 3.2.2-6 70 DRR 15-0944 4/28/15 B 3.2.3-1 0 Amend. No. 123 12/18/99 B 3.2.3-2 0 Amend. No. 123 12/18/99 B 3.2.3-3 0 Amend. No. 123 12/18/99 B 3.2.4-1 0 Amend. No. 123 12/18/99 B 3.2.4-2 0 Amend. No. 123 12/18/99 B 3.2.4-3 48 DRR 10-3740 12/28/10 B 3.2.4-4 0 Amend. No. 123 12/18/99 B 3.2.4-5 48 DRR 10-3740 12/28/10 B 3.2.4-6 0 Amend. No. 123 12/18/99 B 3.2.4-7 48 DRR 10-3740 12/28/10 TAB - B 3.3 INSTRUMENTATION B 3.3.1-1 84 DRR 20-0400 08/18/20 B 3.3.1-2 0 Amend. No. 123 12/18/99 B 3.3.1-3 0 Amend. No. 123 12/18/99 B 3.3.1-4 0 Amend. No. 123 12/18/99 B 3.3.1-5 0 Amend. No. 123 12/18/99 B 3.3.1-6 0 Amend. No. 123 12/18/99 B 3.3.1-7 5 DRR 00-1427 10/12/00 B 3.3.1-8 0 Amend. No. 123 12/18/99 B 3.3.1-9 84 DRR 20-0400 08/18/20 B 3.3.1-10 84 DRR 20-0400 08/18/20 B 3.3.1-11 84 DRR 20-0400 08/18/20 B 3.3.1-12 84 DRR 20-0400 08/18/20 B 3.3.1-13 84 DRR 20-0400 08/18/20 B 3.3.1-14 84 DRR 20-0400 08/18/20 B 3.3.1-15 84 DRR 20-0400 08/18/20 B 3.3.1-16 84 DRR 20-0400 08/18/20 B 3.3.1-17 84 DRR 20-0400 08/18/20 B 3.3.1-18 84 DRR 20-0400 08/18/20 B 3.3.1-19 84 DRR 20-0400 08/18/20 B 3.3.1-20 84 DRR 20-0400 08/18/20 B 3.3.1-21 84 DRR 20-0400 08/18/20 B 3.3.1-22 84 DRR 20-0400 08/18/20 B 3.3.1-23 84 DRR 20-0400 08/18/20 B 3.3.1-24 84 DRR 20-0400 08/18/20 B 3.3.1-25 84 DRR 20-0400 08/18/20 B 3.3.1-26 84 DRR 20-0400 08/18/20 B 3.3.1-27 84 DRR 20-0400 08/18/20 B 3.3.1-28 84 DRR 20-0400 08/18/20 B 3.3.1-29 84 DRR 20-0400 08/18/20 B 3.3.1-30 84 DRR 20-0400 08/18/20 B 3.3.1-31 84 DRR 20-0400 08/18/20 B 3.3.1-32 84 DRR 20-0400 08/18/20 B 3.3.1-33 84 DRR 20-0400 08/18/20 Wolf Creek - Unit 1 iii Revision 86

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.1-34 84 DRR 20-0400 08/18/20 B 3.3.1-35 84 DRR 20-0400 08/18/20 B 3.3.1-36 84 DRR 20-0400 08/18/20 B 3.3.1-37 84 DRR 20-0400 08/18/20 B 3.3.1-38 84 DRR 20-0400 08/18/20 B 3.3.1-39 84 DRR 20-0400 08/18/20 B 3.3.1-40 84 DRR 20-0400 08/18/20 B 3.3.1-41 84 DRR 20-0400 08/18/20 B 3.3.1-42 84 DRR 20-0400 08/18/20 B 3.3.1-43 84 DRR 20-0400 08/18/20 B 3.3.1-44 84 DRR 20-0400 08/18/20 B 3.3.1-45 84 DRR 20-0400 08/18/20 B 3.3.1-46 84 DRR 20-0400 08/18/20 B 3.3.1-47 84 DRR 20-0400 08/18/20 B 3.3.1-48 84 DRR 20-0400 08/18/20 B 3.3.1-49 84 DRR 20-0400 08/18/20 B 3.3.1-50 84 DRR 20-0400 08/18/20 B 3.3.1-51 84 DRR 20-0400 08/18/20 B 3.3.1-52 84 DRR 20-0400 08/18/20 B 3.3.1-53 84 DRR 20-0400 08/18/20 B 3.3.1-54 84 DRR 20-0400 08/18/20 B 3.3.1-55 84 DRR 20-0400 08/18/20 B 3.3.1-56 84 DRR 20-0400 08/18/20 B 3.3.1-57 84 DRR 20-0400 08/18/20 B 3.3.1-58 84 DRR 20-0400 08/18/20 B 3.3.1-59 84 DRR 20-0400 08/18/20 B 3.3.1-60 84 DRR 20-0400 08/18/20 B 3.3.1-61 84 DRR 20-0400 08/18/20 B 3.3.1-62 84 DRR 20-0400 08/18/20 B 3.3.2-1 84 DRR 20-0400 08/18/20 B 3.3.2-2 0 Amend. No. 123 12/18/99 B 3.3.2-3 0 Amend. No. 123 12/18/99 B 3.3.2-4 0 Amend. No. 123 12/18/99 B 3.3.2-5 0 Amend. No. 123 12/18/99 B 3.3.2-6 7 DRR 01-0474 5/1/01 B 3.3.2-7 0 Amend. No. 123 12/18/99 B 3.3.2-8 0 Amend. No. 123 12/18/99 B 3.3.2-9 0 Amend. No. 123 12/18/99 B 3.3.2-10 0 Amend. No. 123 12/18/99 B 3.3.2-11 0 Amend. No. 123 12/18/99 B 3.3.2-12 81 DRR 19-1027 10/28/19 B 3.3.2-13 0 Amend. No. 123 12/18/99 B 3.3.2-14 2 DRR 00-0147 4/24/00 B 3.3.2-15 0 Amend. No. 123 12/18/99 B 3.3.2-16 0 Amend. No. 123 12/18/99 B 3.3.2-17 0 Amend. No. 123 12/18/99 B 3.3.2-18 0 Amend. No. 123 12/18/99 B 3.3.2-19 37 DRR 08-0503 4/8/08 B 3.3.2-20 37 DRR 08-0503 4/8/08 B 3.3.2-21 37 DRR 08-0503 4/8/08 B 3.3.2-22 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 iv Revision 86

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.2-23 37 DRR 08-0503 4/8/08 B 3.3.2-24 39 DRR 08-1096 8/28/08 B 3.3.2-25 39 DRR 08-1096 8/28/08 B 3.3.2-26 39 DRR 08-1096 8/28/08 B 3.3.2-27 37 DRR 08-0503 4/8/08 B 3.3.2-28 84 DRR 20-0400 08/18/20 B 3.3.2-29 0 Amend. No. 123 12/18/99 B 3.3.2-30 0 Amend. No. 123 12/18/99 B 3.3.2-31 52 DRR 11-0724 4/11/11 B 3.3.2-32 52 DRR 11-0724 4/11/11 B 3.3.2-33 0 Amend. No. 123 12/18/99 B 3.3.2-34 0 Amend. No. 123 12/18/99 B 3.3.2-35 20 DRR 04-1533 2/16/05 B 3.3.2-36 20 DRR 04-1533 2/16/05 B 3.3.2-37 20 DRR 04-1533 2/16/05 B 3.3.2-38 20 DRR 04-1533 2/16/05 B 3.3.2-39 25 DRR 06-0800 5/18/06 B 3.3.2-40 20 DRR 04-1533 2/16/05 B 3.3.2-41 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-42 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-43 20 DRR 04-1533 2/16/05 B 3.3.2-44 20 DRR 04-1533 2/16/05 B 3.3.2-45 20 DRR 04-1533 2/16/05 B 3.3.2-46 54 DRR 11-2394 11/16/11 B 3.3.2-47 43 DRR 09-1416 9/2/09 B 3.3.2-48 37 DRR 08-0503 4/8/08 B 3.3.2-49 20 DRR 04-1533 2/16/05 B 3.3.2-50 20 DRR 04-1533 2/16/05 B 3.3.2-51 43 DRR 09-1416 9/2/09 B 3.3.2-52 43 DRR 09-1416 9/2/09 B 3.3.2-53 43 DRR 09-1416 9/2/09 B 3.3.2-54 43 DRR 09-1416 9/2/09 B 3.3.2-55 43 DRR 09-1416 9/2/09 B 3.3.2-56 43 DRR 09-1416 9/2/09 B 3.3.2-57 43 DRR 09-1416 9/2/09 B 3.3.3-1 0 Amend. No. 123 12/18/99 B 3.3.3-2 5 DRR 00-1427 10/12/00 B 3.3.3-3 0 Amend. No. 123 12/18/99 B 3.3.3-4 0 Amend. No. 123 12/18/99 B 3.3.3-5 0 Amend. No. 123 12/18/99 B 3.3.3-6 8 DRR 01-1235 9/19/01 B 3.3.3-7 21 DRR 05-0707 4/20/05 B 3.3.3-8 81 DRR 19-1027 10/28/19 B 3.3.3-9 8 DRR 01-1235 9/19/01 B 3.3.3-10 19 DRR 04-1414 10/12/04 B 3.3.3-11 19 DRR 04-1414 10/12/04 B 3.3.3-12 21 DRR 05-0707 4/20/05 B 3.3.3-13 21 DRR 05-0707 4/20/05 B 3.3.3-14 8 DRR 01-1235 9/19/01 B 3.3.3-15 8 DRR 01-1235 9/19/01 B 3.3.4-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 v Revision 86

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.4-2 9 DRR 02-1023 2/28/02 B 3.3.4-3 15 DRR 03-0860 7/10/03 B 3.3.4-4 19 DRR 04-1414 10/12/04 B 3.3.4-5 1 DRR 99-1624 12/18/99 B 3.3.4-6 9 DRR 02-0123 2/28/02 B 3.3.5-1 0 Amend. No. 123 12/18/99 B 3.3.5-2 1 DRR 99-1624 12/18/99 B 3.3.5-3 1 DRR 99-1624 12/18/99 B 3.3.5-4 1 DRR 99-1624 12/18/99 B 3.3.5-5 0 Amend. No. 123 12/18/99 B 3.3.5-6 22 DRR 05-1375 6/28/05 B 3.3.5-7 22 DRR 05-1375 6/28/05 B 3.3.6-1 81 DRR 19-1027 10/28/19 B 3.3.6-2 81 DRR 19-1027 10/28/19 B 3.3.6-3 0 Amend. No. 123 12/18/99 B 3.3.6-4 0 Amend. No. 123 12/18/99 B 3.3.6-5 0 Amend. No. 123 12/18/99 B 3.3.6-6 0 Amend. No. 123 12/18/99 B 3.3.6-7 81 DRR 19-1027 10/28/19 B 3.3.7-1 81 DRR 19-1027 10/28/19 B 3.3.7-2 81 DRR 19-1027 10/28/19 B 3.3.7-3 57 DRR 13-0006 1/16/13 B 3.3.7-4 0 Amend. No. 123 12/18/99 B 3.3.7-5 0 Amend. No. 123 12/18/99 B 3.3.7-6 57 DRR 13-0006 1/16/13 B 3.3.7-7 0 Amend. No. 123 12/18/99 B 3.3.7-8 81 DRR 19-1027 10/28/19 B 3.3.8-1 84 DRR 20-0400 8/18/20 B 3.3.8-2 0 Amend. No. 123 12/18/99 B 3.3.8-3 57 DRR 13-0006 1/16/13 B 3.3.8-4 57 DRR 13-0006 1/16/13 B 3.3.8-5 0 Amend. No. 123 12/18/99 B 3.3.8-6 24 DRR 06-0051 2/28/06 B 3.3.8-7 81 DRR 19-1027 10/28/19 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1-1 84 DRR 20-0400 08/18/20 B 3.4.1-2 84 DRR 20-0400 08/18/20 B 3.4.1-3 10 DRR 02-0411 4/5/02 B 3.4.1-4 0 Amend. No. 123 12/18/99 B 3.4.1-5 0 Amend. No. 123 12/18/99 B 3.4.1-6 84 DRR 20-0400 08/18/20 B 3.4.2-1 0 Amend. No. 123 12/18/99 B 3.4.2-2 0 Amend. No. 123 12/18/99 B 3.4.2-3 0 Amend. No. 123 12/18/99 B 3.4.3-1 67 DRR 15-0116 2/10/15 B 3.4.3-2 0 Amend. No. 123 12/18/99 B 3.4.3-3 0 Amend. No. 123 12/18/99 B 3.4.3-4 0 Amend. No. 123 12/18/99 B 3.4.3-5 0 Amend. No. 123 12/18/99 B 3.4.3-6 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 vi Revision 86

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.3-7 0 Amend. No. 123 12/18/99 B 3.4.4-1 0 Amend. No. 123 12/18/99 B 3.4.4-2 29 DRR 06-1984 10/17/06 B 3.4.4-3 0 Amend. No. 123 12/18/99 B 3.4.5-1 0 Amend. No. 123 12/18/99 B 3.4.5-2 53 DRR 11-1513 7/18/11 B 3.4.5-3 29 DRR 06-1984 10/17/06 B 3.4.5-4 0 Amend. No. 123 12/18/99 B 3.4.5-5 12 DRR 02-1062 9/26/02 B 3.4.5-6 12 DRR 02-1062 9/26/02 B 3.4.6-1 53 DRR 11-1513 7/18/11 B 3.4.6-2 72 DRR 15-1918 10/26/15 B 3.4.6-3 12 DRR 02-1062 9/26/02 B 3.4.6-4 72 DRR 15-1918 10/26/15 B 3.4.6-5 75 DRR 16-1909 10/26/16 B 3.4.6-6 75 DRR 16-1909 10/26/16 B 3.4.7-1 12 DRR 02-1062 9/26/02 B 3.4.7-2 17 DRR 04-0453 5/26/04 B 3.4.7-3 72 DRR 15-1918 10/26/15 B 3.4.7-4 42 DRR 09-1009 7/16/09 B 3.4.7-5 72 DRR 15-1918 10/26/15 B 3.4.7-6 75 DRR 16-1909 10/26/16 B 3.4.8-1 53 DRR 11-1513 7/18/11 B 3.4.8-2 72 DRR 15-1918 10/26/15 B 3.4.8-3 42 DRR 09-1009 7/16/09 B 3.4.8-4 75 DRR 16-1909 10/26/16 B 3.4.8-5 72 DRR 15-1918 10/26/15 B 3.4.9-1 0 Amend. No. 123 12/18/99 B 3.4.9-2 0 Amend. No. 123 12/18/99 B 3.4.9-3 0 Amend. No. 123 12/18/99 B 3.4.9-4 0 Amend. No. 123 12/18/99 B 3.4.10-1 5 DRR 00-1427 10/12/00 B 3.4.10-2 5 DRR 00-1427 10/12/00 B 3.4.10-3 0 Amend. No. 123 12/18/99 B 3.4.10-4 32 DRR 07-0139 2/7/07 B 3.4.11-1 0 Amend. No. 123 12/18/99 B 3.4.11-2 1 DRR 99-1624 12/18/99 B 3.4.11-3 19 DRR 04-1414 10/12/04 B 3.4.11-4 0 Amend. No. 123 12/18/99 B 3.4.11-5 1 DRR 99-1624 12/18/99 B 3.4.11-6 0 Amend. No. 123 12/18/99 B 3.4.11-7 32 DRR 07-0139 2/7/07 B 3.4.12-1 61 DRR 14-0346 2/27/14 B 3.4.12-2 61 DRR 14-0346 2/27/14 B 3.4.12-3 0 Amend. No. 123 12/18/99 B 3.4.12-4 61 DRR 14-0346 2/27/14 B 3.4.12-5 61 DRR 14-0346 2/27/14 B 3.4.12-6 56 DRR 12-1792 11/7/12 B 3.4.12-7 61 DRR 14-0346 2/27/14 B 3.4.12-8 1 DRR 99-1624 12/18/99 B 3.4.12-9 56 DRR 12-1792 11/7/12 Wolf Creek - Unit 1 vii Revision 86

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IMPLEMENTED (4)

TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.12-10 0 Amend. No. 123 12/18/99 B 3.4.12-11 61 DRR 14-0346 2/27/14 B 3.4.12-12 32 DRR 07-0139 2/7/07 B 3.4.12-13 0 Amend. No. 123 12/18/99 B 3.4.12-14 32 DRR 07-0139 2/7/07 B 3.4.13-1 0 Amend. No. 123 12/18/99 B 3.4.13-2 81 DRR 19-1027 10/28/19 B 3.4.13-3 29 DRR 06-1984 10/17/06 B 3.4.13-4 35 DRR 07-1553 9/28/07 B 3.4.13-5 35 DRR 07-1553 9/28/07 B 3.4.13-6 81 DRR 19-1027 10/28/19 B 3.4.14-1 0 Amend. No. 123 12/18/99 B 3.4.14-2 0 Amend. No. 123 12/18/99 B 3.4.14-3 0 Amend. No. 123 12/18/99 B 3.4.14-4 0 Amend. No. 123 12/18/99 B 3.4.14-5 32 DRR 07-0139 2/7/07 B 3.4.14-6 32 DRR 07-0139 2/7/07 B 3.4.15-1 31 DRR 06-2494 12/13/06 B 3.4.15-2 31 DRR 06-2494 12/13/06 B 3.4.15-3 33 DRR 07-0656 5/1/07 B 3.4.15-4 33 DRR 07-0656 5/1/07 B 3.4.15-5 65 DRR 14-2146 9/30/14 B 3.4.15-6 31 DRR 06-2494 12/13/06 B 3.4.15-7 31 DRR 06-2494 12/13/06 B 3.4.15-8 31 DRR 06-2494 12/13/06 B 3.4.16-1 81 DRR 19-1027 10/28/19 B 3.4.16-2 84 DRR 20-0400 08/18/20 B 3.4.16-3 31 DRR 06-2494 12/13/06 B 3.4.16-4 31 DRR 06-2494 12/13/06 B 3.4.16-5 81 DRR 19-1027 10/28/19 B 3.4.17-1 29 DRR 06-1984 10/17/06 B 3.4.17-2 81 DRR 19-1027 10/28/19 B 3.4.17-3 52 DRR 11-0724 4/11/11 B 3.4.17-4 81 DRR 19-1027 10/28/19 B 3.4.17-5 57 DRR 13-0006 1/16/13 B 3.4.17-6 57 DRR 13-0006 1/16/13 B 3.4.17-7 81 DRR 19-1027 10/28/19 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1-1 0 Amend. No. 123 12/18/99 B 3.5.1-2 0 Amend. No. 123 12/18/99 B 3.5.1-3 73 DRR 15-2135 11/17/15 B 3.5.1-4 73 DRR 15-2135 11/17/15 B 3.5.1-5 1 DRR 99-1624 12/18/99 B 3.5.1-6 1 DRR 99-1624 12/18/99 B 3.5.1-7 71 DRR 15-1528 7/30/15 B 3.5.1-8 1 DRR 99-1624 12/18/99 B 3.5.2-1 84 DRR 20-0400 08/18/20 B 3.5.2-2 0 Amend. No. 123 12/18/99 B 3.5.2-3 0 Amend. No. 123 12/18/99 B 3.5.2-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 viii Revision 86

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TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

B 3.5.2-5 72 DRR 15-1918 10/26/15 B 3.5.2-6 42 DRR 09-1009 7/16/09 B 3.5.2-7 42 DRR 09-1009 7/16/09 B 3.5.2-8 72 DRR 15-1918 10/26/15 B 3.5.2-9 75 DRR 16-1909 10/26/16 B 3.5.2-10 80 DRR 19-0524 5/30/19 B 3.5.2-11 72 DRR 15-1918 10/26/15 B 3.5.2-12 72 DRR 15-1918 10/26/15 B 3.5.3-1 56 DRR 12-1792 11/7/12 B 3.5.3-2 72 DRR 15-1918 10/26/15 B 3.5.3-3 56 DRR 12-1792 11/7/12 B 3.5.3-4 56 DRR 12-1792 11/7/12 B 3.5.4-1 0 Amend. No. 123 12/18/99 B 3.5.4-2 0 Amend. No. 123 12/18/99 B 3.5.4-3 0 Amend. No. 123 12/18/99 B 3.5.4-4 0 Amend. No. 123 12/18/99 B 3.5.4-5 0 Amend. No. 123 12/18/99 B 3.5.4-6 26 DRR 06-1350 7/24/06 B 3.5.5-1 21 DRR 05-0707 4/20/05 B 3.5.5-2 21 DRR 05-0707 4/20/05 B 3.5.5-3 2 Amend. No. 132 4/24/00 B 3.5.5-4 21 DRR 05-0707 4/20/05 TAB - B 3.6 CONTAINMENT SYSTEMS B 3.6.1-1 0 Amend. No. 123 12/18/99 B 3.6.1-2 81 DRR 19-1027 10/28/19 B 3.6.1-3 0 Amend. No. 123 12/18/99 B 3.6.1-4 17 DRR 04-0453 5/26/04 B 3.6.2-1 81 DRR 19-1027 10/28/19 B 3.6.2-2 0 Amend. No. 123 12/18/99 B 3.6.2-3 0 Amend. No. 123 12/18/99 B 3.6.2-4 0 Amend. No. 123 12/18/99 B 3.6.2-5 0 Amend. No. 123 12/18/99 B 3.6.2-6 0 Amend. No. 123 12/18/99 B 3.6.2-7 0 Amend. No. 123 12/18/99 B 3.6.3-1 0 Amend. No. 123 12/18/99 B 3.6.3-2 84 DRR 20-0400 08/18/20 B 3.6.3-3 81 DRR 19-1027 10/28/19 B 3.6.3-4 49 DRR 11-0014 1/31/11 B 3.6.3-5 49 DRR 11-0014 1/31/11 B 3.6.3-6 49 DRR 11-0014 1/31/11 B 3.6.3-7 41 DRR 09-0288 3/20/09 B 3.6.3-8 36 DRR 08-0255 3/11/08 B 3.6.3-9 36 DRR 08-0255 3/11/08 B 3.6.3-10 8 DRR 01-1235 9/19/01 B 3.6.3-11 36 DRR 08-0255 3/11/08 B 3.6.3-12 36 DRR 08-0255 3/11/08 B 3.6.3-13 50 DRR 11-0449 3/9/11 B 3.6.3-14 36 DRR 08-0255 3/11/08 B 3.6.3-15 39 DRR 08-1096 8/28/08 B 3.6.3-16 39 DRR 08-1096 8/28/08 Wolf Creek - Unit 1 ix Revision 86

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TAB - B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.3-17 36 DRR 08-0255 3/11/08 B 3.6.3-18 36 DRR 08-0255 3/11/08 B 3.6.3-19 36 DRR 08-0255 3/11/08 B 3.6.4-1 39 DRR 08-1096 8/28/08 B 3.6.4-2 0 Amend. No. 123 12/18/99 B 3.6.4-3 0 Amend. No. 123 12/18/99 B 3.6.5-1 0 Amend. No. 123 12/18/99 B 3.6.5-2 37 DRR 08-0503 4/8/08 B 3.6.5-3 13 DRR 02-1458 12/03/02 B 3.6.5-4 0 Amend. No. 123 12/18/99 B 3.6.6-1 81 DRR 19-1027 10/28/19 B 3.6.6-2 63 DRR 14-1572 7/1/14 B 3.6.6-3 37 DRR 08-0503 4/8/08 B 3.6.6-4 81 DRR 19-1027 10/28/19 B 3.6.6-5 0 Amend. No. 123 12/18/99 B 3.6.6-6 18 DRR 04-1018 9/1/04 B 3.6.6-7 72 DRR 15-1918 10/26/15 B 3.6.6-8 80 DRR 19-0524 5/30/19 B 3.6.6-9 72 DRR 15-1918 10/26/15 B 3.6.6-10 75 DRR 16-1909 10/26/16 B 3.6.6.11 80 DRR 19-0524 5/30/19 B 3.6.7-1 0 Amend. No. 123 12/18/99 B 3.6.7-2 81 DRR 19-1027 10/28/19 B 3.6.7-3 81 DRR 19-1027 10/28/19 B 3.6.7-4 81 DRR 19-1027 10/28/19 B 3.6.7-5 42 DRR 09-1009 7/16/09 TAB - B 3.7 PLANT SYSTEMS B 3.7.1-1 0 Amend. No. 123 12/18/99 B 3.7.1-2 84 DRR 20-0400 08/18/20 B 3.7.1-3 0 Amend. No. 123 12/18/99 B 3.7.1-4 84 DRR 20-0400 08/18/20 B 3.7.1-5 84 DRR 20-0400 08/18/20 B 3.7.1-6 84 DRR 20-0400 08/18/20 B 3.7.2-1 44 DRR 09-1744 10/28/09 B 3.7.2-2 44 DRR 09-1744 10/28/09 B 3.7.2-3 44 DRR 09-1744 10/28/09 B 3.7.2-4 81 DRR 19-1027 10/28/19 B 3.7.2-5 44 DRR 09-1744 10/28/09 B 3.7.2-6 44 DRR 09-1744 10/28/09 B 3.7.2-7 44 DRR 09-1744 10/28/09 B 3.7.2-8 44 DRR 09-1744 10/28/09 B 3.7.2-9 44 DRR 09-1744 10/28/09 B 3.7.2-10 81 DRR 19-1027 10/28/19 B 3.7.2-11 44 DRR 09-1744 10/28/09 B 3.7.3-1 37 DRR 08-0503 4/8/08 B 3.7.3-2 50 DRR 11-0449 3/9/11 B 3.7.3-3 37 DRR 08-0503 4/8/08 B 3.7.3-4 37 DRR 08-0503 4/8/08 B 3.7.3-5 37 DRR 08-0503 4/8/08 B 3.7.3-6 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 x Revision 86

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.3-7 37 DRR 08-0503 4/8/08 B 3.7.3-8 37 DRR 08-0503 4/8/08 B 3.7.3-9 66 DRR 14-2329 11/6/14 B 3.7.3-10 66 DRR 14-2329 11/6/14 B 3.7.3-11 37 DRR 08-0503 4/8/08 B 3.7.4-1 1 DRR 99-1624 12/18/99 B 3.7.4-2 84 DRR 20-0400 08/18/20 B 3.7.4-3 19 DRR 04-1414 10/12/04 B 3.7.4-4 19 DRR 04-1414 10/12/04 B 3.7.4-5 84 DRR 20-0400 08/18/20 B 3.7.5-1 54 DRR 11-2394 11/16/11 B 3.7.5-2 54 DRR 11-2394 11/16/11 B 3.7.5-3 0 Amend. No. 123 12/18/99 B 3.7.5-4 85 DRR 20-0988 10/24/20 B 3.7.5-5 76 DRR 17-0343 2/21/17 B 3.7.5-6 85 DRR 20-0988 10/24/20 B 3.7.5-7 85 DRR 20-0988 10/24/20 B 3.7.5-8 85 DRR 20-0988 10/24/20 B 3.7.5-9 85 DRR 20-0988 10/24/20 B 3.7.5-10 85 DRR 20-0988 10/24/20 B 3.7.6-1 0 Amend. No. 123 12/18/99 B 3.7.6-2 0 Amend. No. 123 12/18/99 B 3.7.6-3 0 Amend. No. 123 12/18/99 B 3.7.7-1 0 Amend. No. 123 12/18/99 B 3.7.7-2 0 Amend. No. 123 12/18/99 B 3.7.7-3 0 Amend. No. 123 12/18/99 B 3.7.7-4 1 DRR 99-1624 12/18/99 B 3.7.8-1 0 Amend. No. 123 12/18/99 B 3.7.8-2 0 Amend. No. 123 12/18/99 B 3.7.8-3 0 Amend. No. 123 12/18/99 B 3.7.8-4 0 Amend. No. 123 12/18/99 B 3.7.8-5 0 Amend. No. 123 12/18/99 B 3.7.9-1 3 Amend. No. 134 7/14/00 B 3.7.9-2 3 Amend. No. 134 7/14/00 B 3.7.9-3 3 Amend. No. 134 7/14/00 B 3.7.9-4 3 Amend. No. 134 7/14/00 B 3.7.10-1 64 DRR 14-1822 8/28/14 B 3.7.10-2 81 DRR 19-1027 10/28/19 B 3.7.10-3 81 DRR 19-1027 10/28/19 B 3.7.10-4 81 DRR 19-1027 10/28/19 B 3.7.10-5 81 DRR 19-1027 10/28/19 B 3.7.10-6 57 DRR 13-0006 1/16/13 B 3.7.10-7 64 DRR 14-1822 8/28/14 B 3.7.10-8 81 DRR 19-1027 10/28/19 B 3.7.10-9 81 DRR 19-1027 10/28/19 B 3.7.11-1 0 Amend. No. 123 12/18/99 B 3.7.11-2 57 DRR 13-0006 1/16/13 B 3.7.11-3 63 DRR 14-1572 7/1/14 B 3.7.11-4 63 DRR 14-1572 7/1/14 B 3.7.12-1 0 Amend. No. 123 12/18/99 B 3.7.13-1 24 DRR 06-0051 2/28/06 Wolf Creek - Unit 1 xi Revision 86

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.13-2 81 DRR 19-1027 10/28/19 B 3.7.13-3 81 DRR 19-1027 10/28/19 B 3.7.13-4 81 DRR 19-1027 10/28/19 B 3.7.13-5 81 DRR 19-1027 10/28/19 B 3.7.13-6 81 DRR 19-1027 10/28/19 B 3.7.13-7 81 DRR 19-1027 10/28/19 B 3.7.13-8 81 DRR 19-1027 10/28/19 B 3.7.14-1 0 Amend. No. 123 12/18/99 B 3.7.15-1 81 DRR 19-1027 10/28/19 B 3.7.15-2 81 DRR 19-1027 10/28/19 B 3.7.15-3 81 DRR 19-1027 10/28/19 B 3.7.16-1 5 DRR 00-1427 10/12/00 B 3.7.16-2 23 DRR 05-1995 9/28/05 B 3.7.16-3 5 DRR 00-1427 10/12/00 B 3.7.17-1 7 DRR 01-0474 5/1/01 B 3.7.17-2 7 DRR 01-0474 5/1/01 B 3.7.17-3 5 DRR 00-1427 10/12/00 B 3.7.18-1 81 DRR 19-1027 10/28/19 B 3.7.18-2 81 DRR 19-1027 10/28/19 B 3.7.18-3 81 DRR 19-1027 10/28/19 B 3.7.19-1 44 DRR 09-1744 10/28/09 B 3.7.19-2 54 DRR 11-2394 11/16/11 B 3.7.19-3 54 DRR 11-2394 11/16/11 B 3.7.19-4 61 DRR 14-0346 2/27/14 B 3.7.19-5 61 DRR 14-0346 2/27/14 B 3.7.19-6 54 DRR 11-2394 11/16/11 B 3.7.19-7 54 DRR 11-2394 11/16/11 B 3.7.20-1 79 DRR 18-1579 10/22/18 B 3.7.20-2 79 DRR 18-1579 10/22/18 B 3.7.20-3 85 DRR 20-0988 10/24/20 B 3.7.20-4 79 DRR 18-1579 10/22/18 B 3.7.20-5 79 DRR 18-1579 10/22/18 TAB - B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1-1 54 DRR 11-2394 11/16/11 B 3.8.1-2 0 Amend. No. 123 12/18/99 B 3.8.1-3 86 DRR 21-0212 2/18/21 B 3.8.1-4 86 DRR 21-0212 2/18/21 B 3.8.1-5 59 DRR 13-1524 6/26/13 B 3.8.1-6 25 DRR 06-0800 5/18/06 B 3.8.1-7 26 DRR 06-1350 7/24/06 B 3.8.1-8 35 DRR 07-1553 9/28/07 B 3.8.1-9 42 DRR 09-1009 7/16/09 B 3.8.1-10 39 DRR 08-1096 8/28/08 B 3.8.1-11 36 DRR 08-0255 3/11/08 B 3.8.1-12 75 DRR 16-1909 10/26/16 B 3.8.1-13 47 DRR 10-1089 6/16/10 B 3.8.1-14 47 DRR 10-1089 6/16/10 B 3.8.1-15 47 DRR 10-1089 6/16/10 B 3.8.1-16 26 DRR 06-1350 7/24/06 B 3.8.1-17 26 DRR 06-1350 7/24/06 Wolf Creek - Unit 1 xii Revision 86

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.1-18 59 DRR 13-1524 6/26/13 B 3.8.1-19 26 DRR 06-1350 7/24/06 B 3.8.1-20 26 DRR 06-1350 7/24/06 B 3.8.1-21 33 DRR 07-0656 5/1/07 B 3.8.1-22 33 DRR 07-0656 5/1/07 B 3.8.1-23 74 DRR 16-1182 7/7/16 B 3.8.1-24 74 DRR 16-1182 7/7/16 B 3.8.1-25 74 DRR 16-1182 7/7/16 B 3.8.1-26 74 DRR 16-1182 7/7/16 B 3.8.1-27 74 DRR 16-1182 7/7/16 B 3.8.1-28 74 DRR 16-1182 7/7/16 B 3.8.1-29 74 DRR 16-1182 7/7/16 B 3.8.1-30 74 DRR 16-1182 7/7/16 B 3.8.1-31 74 DRR 16-1182 7/7/16 B 3.8.1-32 74 DRR 16-1182 7/7/16 B 3.8.1-33 74 DRR 16-1182 7/7/16 B 3.8.1-34 74 DRR 16-1182 7/7/16 B 3.8.2-1 57 DRR 13-0006 1/16/13 B 3.8.2-2 0 Amend. No. 123 12/18/99 B 3.8.2-3 86 DRR 21-0212 2/18/21 B 3.8.2-4 86 DRR 21-0212 2/18/21 B 3.8.2-5 57 DRR 13-0006 1/16/13 B 3.8.2-6 57 DRR 13-0006 1/16/13 B 3.8.2-7 57 DRR 13-0006 1/16/13 B 3.8.3-1 1 DRR 99-1624 12/18/99 B 3.8.3-2 0 Amend. No. 123 12/18/99 B 3.8.3-3 0 Amend. No. 123 12/18/99 B 3.8.3-4 1 DRR 99-1624 12/18/99 B 3.8.3-5 0 Amend. No. 123 12/18/99 B 3.8.3-6 0 Amend. No. 123 12/18/99 B 3.8.3-7 12 DRR 02-1062 9/26/02 B 3.8.3-8 1 DRR 99-1624 12/18/99 B 3.8.3-9 0 Amend. No. 123 12/18/99 B 3.8.4-1 0 Amend. No. 123 12/18/99 B 3.8.4-2 0 Amend. No. 123 12/18/99 B 3.8.4-3 0 Amend. No. 123 12/18/99 B 3.8.4-4 0 Amend. No. 123 12/18/99 B 3.8.4-5 50 DRR 11-0449 3/9/11 B 3.8.4-6 50 DRR 11-0449 3/9/11 B 3.8.4-7 6 DRR 00-1541 3/13/01 B 3.8.4-8 0 Amend. No. 123 12/18/99 B 3.8.4-9 2 DRR 00-0147 4/24/00 B 3.8.5-1 57 DRR 13-0006 1/16/13 B 3.8.5-2 0 Amend. No. 123 12/18/99 B 3.8.5-3 57 DRR 13-0006 1/16/13 B 3.8.5-4 57 DRR 13-0006 1/16/13 B 3.8.5-5 57 DRR 13-0006 1/16/13 B 3.8.6-1 0 Amend. No. 123 12/18/99 B 3.8.6-2 0 Amend. No. 123 12/18/99 B 3.8.6-3 0 Amend. No. 123 12/18/99 B 3.8.6-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiii Revision 86

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.6-5 0 Amend. No. 123 12/18/99 B 3.8.6-6 0 Amend. No. 123 12/18/99 B 3.8.7-1 69 DRR 15-0493 3/26/15 B 3.8.7-2 69 DRR 15-0493 3/26/15 B 3.8.7-3 69 DRR 15-0493 3/26/15 B 3.8.7-4 0 Amend. No. 123 12/18/99 B 3.8.8-1 57 DRR 13-0006 1/16/13 B 3.8.8-2 0 Amend. No. 123 12/18/99 B 3.8.8-3 69 DRR 15-0493 3/26/15 B 3.8.8-4 57 DRR 13-0006 1/16/13 B 3.8.8-5 69 DRR 15-0493 3/26/15 B 3.8.9-1 54 DRR 11-2394 11/16/11 B 3.8.9-2 69 DRR 15-0493 3/26/15 B 3.8.9-3 54 DRR 11-2394 11/16/11 B 3.8.9-4 0 Amend. No. 123 12/18/99 B 3.8.9-5 69 DRR 15-0493 3/26/15 B 3.8.9-6 0 Amend. No. 123 12/18/99 B 3.8.9-7 0 Amend. No. 123 12/18/99 B 3.8.9-8 1 DRR 99-1624 12/18/99 B 3.8.9-9 0 Amend. No. 123 12/18/99 B 3.8.10-1 57 DRR 13-0006 1/16/13 B 3.8.10-2 0 Amend. No. 123 12/18/99 B 3.8.10-3 0 Amend. No. 123 12/18/99 B 3.8.10-4 57 DRR 13-0006 1/16/13 B 3.8.10-5 57 DRR 13-0006 1/16/13 B 3.8.10-6 57 DRR 13-0006 1/16/13 TAB - B 3.9 REFUELING OPERATIONS B 3.9.1-1 0 Amend. No. 123 12/18/99 B 3.9.1-2 19 DRR 04-1414 10/12/04 B 3.9.1-3 19 DRR 04-1414 10/12/04 B 3.9.1-4 19 DRR 04-1414 10/12/04 B 3.9.2-1 0 Amend. No. 123 12/18/99 B 3.9.2-2 0 Amend. No. 123 12/18/99 B 3.9.2-3 0 Amend. No. 123 12/18/99 B 3.9.3-1 68 DRR 15-0248 2/26/15 B 3.9.3-2 68 DRR 15-0248 2/26/15 B 3.9.3-3 51 DRR 11-0664 3/21/11 B 3.9.3-4 68 DRR 15-0248 2/26/15 B 3.9.4-1 81 DRR 19-1027 10/28/19 B 3.9.4-2 13 DRR 02-1458 12/03/02 B 3.9.4-3 81 DRR 19-1027 10/28/19 B 3.9.4-4 23 DRR 05-1995 9/28/05 B 3.9.4-5 33 DRR 07-0656 5/1/07 B 3.9.4-6 23 DRR 05-1995 9/28/05 B 3.9.5-1 0 Amend. No. 123 12/18/99 B 3.9.5-2 72 DRR 15-1918 10/26/15 B 3.9.5-3 32 DRR 07-0139 2/7/07 B 3.9.5-4 75 DRR 16-1909 10/26/16 B 3.9.5-5 75 DRR 16-1909 10/26/16 B 3.9.6-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiv Revision 86

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IMPLEMENTED (4)

TAB - B 3.9 REFUELING OPERATIONS (continued)

B 3.9.6-2 72 DRR 15-1918 10/26/15 B 3.9.6-3 42 DRR 09-1009 7/16/09 B 3.9.6-4 72 DRR 15-1918 10/26/15 B 3.9.6-5 75 DRR 16-1909 10/26/16 B 3.9.7-1 81 DRR 19-1027 10/28/19 B 3.9.7-2 81 DRR 19-1027 10/28/19 B 3.9.7-3 81 DRR 19-1027 10/28/19 Note 1 The page number is listed on the center of the bottom of each page.

Note 2 The revision number is listed in the lower right hand corner of each page. The Revision number will be page specific.

Note 3 The change document will be the document requesting the change. Amendment No.

123 issued the improved Technical Specifications and associated Bases which affected each page. The NRC has indicated that Bases changes will not be issued with License Amendments. Therefore, the change document should be a DRR number in accordance with AP 26A-002.

Note 4 The date effective or implemented is the date the Bases pages are issued by Document Control.

Wolf Creek - Unit 1 xv Revision 86

ENCLOSURE Page 16 of 264 AC Sources - Operating B 3.8.1 BASES APPLICABLE meeting the design basis of the unit. This results in maintaining at least SAFETY ANALYSES one train of the onsite or offsite AC sources OPERABLE during Accident (continued) conditions in the event of:

a. An assumed loss of all offsite power or all onsite AC power; and

b. A worst case single failure.

The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two qualified circuits between the offsite transmission network and the onsite Class 1E Electrical Power System, separate and independent DGs for each train, and redundant LSELS for each train ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (AOO) or a postulated DBA.

Each offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while connected to the ESF buses.

One offsite circuit consists of the #7 transformer feeding through the 13-48 breaker power the ESF transformer XNB01, which, in turn powers the NB01 bus through its normal feeder breaker. Transformer XNB01 may also be powered from the SL-7 supply through the 13-8 breaker provided the offsite 69 KV line is not connected to the 345 kV system.

The lineup for transformers 4, 5, and 6 is an alternative to the #7 transformer. However, the alternate lineup does not have open phase detection instrumentation. The lineup with the 4, 5, and 6 transformers should only be used as an alternative in a risk analyzed condition and is not intended to be used for convenience of maintenance. The #7 transformer lineup should be restored as quickly as possible. The offsite circuit energizing NB01 is considered inoperable when NB01 is only energized from the transmission network through the 345-50 and 345-60 main generator breakers. For this configuration, switchyard breakers 345-120 and 345-80 are open.

Another offsite circuit consists of the startup transformer feeding through breaker PA201 powering the ESF transformer XNB02, which, in turn powers the NB02 bus through its normal feeder breaker.

Each DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage. This will be accomplished within 12 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be Wolf Creek - Unit 1 B 3.8.1-3 Revision 86

ENCLOSURE Page 17 of 264 AC Sources - Operating B 3.8.1 BASES LCO met from a variety of initial conditions such as DG in standby with the (continued) engine hot and DG in standby with the engine at ambient conditions.

Additional DG capabilities must be demonstrated to meet required Surveillance, e.g., capability of the DG to revert to standby status on an ECCS signal while operating in parallel test mode.

Upon failure of the DG lube oil keep warm system when the DG is in the standby condition, the DG remains OPERABLE if lube oil temperature is 115 °F and engine lubrication (i.e., flow of lube oil to the DG engine) is maintained. Upon failure of the DG jacket water keep warm system, the DG remains OPERABLE as long as jacket water temperature is 105 F (Ref. 13).

Initiating an EDG start upon a detected undervoltage or degraded voltage condition, tripping of nonessential loads, and proper sequencing of loads, is a required function of LSELS and required for DG OPERABILITY. In addition, the LSELS Automatic Test Indicator (ATI) is an installed testing aid and is not required to be OPERABLE to support the sequencer function. Absence of a functioning ATI does not render LSELS inoperable.

The AC sources in one train must be separate and independent of the AC sources in the other train. For the DGs, separation and independence are complete. For the offsite AC source, separation and independence are to the extent practical.

APPLICABILITY The AC sources and LSELS are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:

a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and

b. Adequate core cooling is provided and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.

The AC power requirements for MODES 5 and 6 are covered in LCO 3.8.2, "AC Sources - Shutdown."

ACTIONS A Note prohibits the application of LCO 3.0.4b. to an inoperable DG.

There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable DG and the provisions of LCO 3.0.4b., 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.

Wolf Creek - Unit 1 B 3.8.1-4 Revision 86

ENCLOSURE Page 18 of 264 AC Sources - Shutdown B 3.8.2 BASES APPLICABLE The AC sources satisfy Criterion 3 of the 10 CFR 50.36(c)(2)(ii).

SAFETY ANALYSES (continued)

LCO One offsite circuit capable of supplying the onsite Class 1E power distribution subsystem of LCO 3.8.10, "Distribution Systems - Shutdown,"

ensures that one train of required loads are powered from offsite power.

An OPERABLE DG, associated with the distribution system train required to be OPERABLE by LCO 3.8.10, ensures a diverse power source is available to provide electrical power support, assuming a loss of the offsite circuit. Together, OPERABILITY of the required offsite circuit and DG ensures the availability of sufficient AC sources to operate the unit in a safe manner and to mitigate the consequences of postulated events during shutdown (e.g., fuel handling accidents).

The DG must be supporting the train of AC electrical distribution required to be OPERABLE per LCO 3.8.10. The offsite circuit must also support the train of AC electrical distribution required to be OPERABLE per LCO 3.8.10. When the second AC electrical power distribution train (subsystem) is needed to support redundant required systems, equipment and components, the second train may be energized from any available source. The available source must be Class 1E or another reliable source. The available source must be capable of supplying sufficient AC electrical power such that the redundant components are capable of performing their specified safety function(s) (implicitly required by the definition of OPERABILITY). Otherwise, the supported components must be declared inoperable and the appropriate conditions of the LCOs for the redundant components must be entered.

The qualified offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while connected to the Engineered Safety Feature (ESF) bus(es).

Qualified offsite circuits are those that are described in the USAR and are part of the licensing basis for the unit.

One offsite circuit consists of the #7 transformer feeding through the 13-48 breaker power the ESF transformer XNB01, which, in turn powers the NB01 bus through its normal feeder breaker. Transformer XNB01 may also be powered from the SL-7 supply through the 13-8 breaker provided that no offsite load is connected to the 69 KV system. The lineup for transformers 4, 5, and 6 is an alternative to the #7 transformer.

However, the alternate lineup does not have open phase detection instrumentation. The lineup with the 4, 5, and 6 transformers should only be used as an alternative in a risk analyzed condition and is not intended to be used for convenience of maintenance. The #7 transformer lineup should be restored as quickly as possible. Another offsite circuit consists Wolf Creek - Unit 1 B 3.8.2-3 Revision 86

ENCLOSURE Page 19 of 264 AC Sources - Shutdown B 3.8.2 BASES LCO of the startup transformer feeding through breaker PA201 powering the (continued) ESF transformer XNB02, which in turn powers the NB02 bus through its normal feeder breaker.

The DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage. This sequence must be accomplished within 12 seconds.

The DG must be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions such as DG in standby with the engine hot and DG in standby at ambient conditions.

Initiating an EDG start upon a detected under voltage or degraded voltage condition, tripping of nonessential loads, and proper sequencing of loads, is a required function of load shedder and emergency load sequencer (LSELS) and required for DG OPERABILITY. Only the shutdown sequencer on the train supported by the OPERABLE DG is required to be OPERABLE in MODES 5 and 6. In addition, the LSELS Automatic Test Indicator (ATI) is an installed testing aid and is not required to be OPERABLE to support the sequencer function. Absence of a functioning ATI does not render LSELS inoperable.

It is acceptable for one offsite circuit to be connected to more than one ESF bus through the normal or alternate feeder breakers with the loading limitations of calculation XX-E-006, "AC System Analysis," not being exceeded. Each offsite circuit can be manually aligned to supply power to the opposite NB bus.

APPLICABILITY The AC sources required to be OPERABLE in MODES 5 and 6, and during movement of irradiated fuel assemblies provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC power requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.1.

Wolf Creek - Unit 1 B 3.8.2-4 Revision 86

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TAB - Title Page Technical Specification Cover Page Title Page TAB - Table of Contents i 34 DRR 07-1057 7/10/07 ii 29 DRR 06-1984 10/17/06 iii 44 DRR 09-1744 10/28/09 TAB - B 2.0 SAFETY LIMITS (SLs)

B 2.1.1-1 0 Amend. No. 123 12/18/99 B 2.1.1-2 14 DRR 03-0102 2/12/03 B 2.1.1-3 14 DRR 03-0102 2/12/03 B 2.1.1-4 0 Amend. No. 123 2/12/03 B 2.1.2-1 84 DRR 20-0400 08/18/20 B 2.1.2-2 84 DRR 20-0400 08/18/20 B 2.1.2-3 81 DRR 19-1027 10/28/19 TAB - B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 81 DRR 19-1027 10/28/19 B 3.0-2 0 Amend. No. 123 12/18/99 B 3.0-3 81 DRR 19-1027 10/28/19 B 3.0-4 81 DRR 19-1027 10/28/19 B 3.0-5 81 DRR 19-1027 10/28/19 B 3.0-6 81 DRR 19-1027 10/28/19 B 3.0-7 81 DRR 19-1027 10/28/19 B 3.0-8 81 DRR 19-1027 10/28/19 B 3.0-9 81 DRR 19-1027 10/28/19 B 3.0-10 81 DRR 19-1027 10/28/19 B 3.0-11 81 DRR 19-1027 10/28/19 B 3.0-12 81 DRR 19-1027 10/28/19 B 3.0-13 81 DRR 19-1027 10/28/19 B 3.0-14 81 DRR 19-1027 10/28/19 B 3.0-15 81 DRR 19-1027 10/28/19 B 3.0-16 81 DRR 19-1027 10/28/19 B 3.0-17 81 DRR 19-1027 10/28/19 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1-1 0 Amend. No. 123 12/18/99 B 3.1.1-2 0 Amend. No. 123 12/18/99 B 3.1.1-3 0 Amend. No. 123 12/18/99 B 3.1.1-4 81 DRR 19-1027 10/28/19 B 3.1.1-5 81 DRR 19-1027 10/28/19 B 3.1.2-1 0 Amend. No. 123 12/18/99 B 3.1.2-2 0 Amend. No. 123 12/18/99 B 3.1.2-3 0 Amend. No. 123 12/18/99 B 3.1.2-4 0 Amend. No. 123 12/18/99 B 3.1.2-5 0 Amend. No. 123 12/18/99 B 3.1.3-1 0 Amend. No. 123 12/18/99 B 3.1.3-2 0 Amend. No. 123 12/18/99 B 3.1.3-3 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 i Revision 87

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TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)

B 3.1.3-4 0 Amend. No. 123 12/18/99 B 3.1.3-5 0 Amend. No. 123 12/18/99 B 3.1.3-6 0 Amend. No. 123 12/18/99 B 3.1.4-1 0 Amend. No. 123 12/18/99 B 3.1.4-2 0 Amend. No. 123 12/18/99 B 3.1.4-3 48 DRR 10-3740 12/28/10 B 3.1.4-4 0 Amend. No. 123 12/18/99 B 3.1.4-5 0 Amend. No. 123 12/18/99 B 3.1.4-6 48 DRR 10-3740 12/28/10 B 3.1.4-7 0 Amend. No. 123 12/18/99 B 3.1.4-8 0 Amend. No. 123 12/18/99 B 3.1.4-9 0 Amend. No. 123 12/18/99 B 3.1.5-1 0 Amend. No. 123 12/18/99 B 3.1.5-2 0 Amend. No. 123 12/18/99 B 3.1.5-3 0 Amend. No. 123 12/18/99 B 3.1.5-4 0 Amend. No. 123 12/18/99 B 3.1.6-1 0 Amend. No. 123 12/18/99 B 3.1.6-2 0 Amend. No. 123 12/18/99 B 3.1.6-3 0 Amend. No. 123 12/18/99 B 3.1.6-4 0 Amend. No. 123 12/18/99 B 3.1.6-5 0 Amend. No. 123 12/18/99 B 3.1.6-6 0 Amend. No. 123 12/18/99 B 3.1.7-1 0 Amend. No. 123 12/18/99 B 3.1.7-2 0 Amend. No. 123 12/18/99 B 3.1.7-3 48 DRR 10-3740 12/28/10 B 3.1.7-4 48 DRR 10-3740 12/28/10 B 3.1.7-5 48 DRR 10-3740 12/28/10 B 3.1.7-6 0 Amend. No. 123 12/18/99 B 3.1.8-1 0 Amend. No. 123 12/18/99 B 3.1.8-2 0 Amend. No. 123 12/18/99 B 3.1.8-3 15 DRR 03-0860 7/10/03 B 3.1.8-4 15 DRR 03-0860 7/10/03 B 3.1.8-5 0 Amend. No. 123 12/18/99 B 3.1.8-6 5 DRR 00-1427 10/12/00 B 3.1.9-1 84 DRR 20-0400 08/18/20 B 3.1.9-2 84 DRR 20-0400 08/18/20 B 3.1.9-3 84 DRR 20-0400 08/18/20 B 3.1.9-4 84 DRR 20-0400 08/18/20 B 3.1.9-5 84 DRR 20-0400 08/18/20 TAB - B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1-1 48 DRR 10-3740 12/28/10 B 3.2.1-2 0 Amend. No. 123 12/18/99 B 3.2.1-3 48 DRR 10-3740 12/28/10 B 3.2.1-4 48 DRR 10-3740 12/28/10 B 3.2.1-5 48 DRR 10-3740 12/28/10 B 3.2.1-6 48 DRR 10-3740 12/28/10 B 3.2.1-7 48 DRR 10-3740 12/28/10 B 3.2.1-8 48 DRR 10-3740 12/28/10 B 3.2.1-9 29 DRR 06-1984 10/17/06 B 3.2.1-10 70 DRR 15-0944 4/28/15 Wolf Creek - Unit 1 ii Revision 87

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TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued)

B 3.2.2-1 48 DRR 10-3740 12/28/10 B 3.2.2-2 0 Amend. No. 123 12/18/99 B 3.2.2-3 48 DRR 10-3740 12/28/10 B 3.2.2-4 48 DRR 10-3740 12/28/10 B 3.2.2-5 48 DRR 10-3740 12/28/10 B 3.2.2-6 70 DRR 15-0944 4/28/15 B 3.2.3-1 0 Amend. No. 123 12/18/99 B 3.2.3-2 0 Amend. No. 123 12/18/99 B 3.2.3-3 0 Amend. No. 123 12/18/99 B 3.2.4-1 0 Amend. No. 123 12/18/99 B 3.2.4-2 0 Amend. No. 123 12/18/99 B 3.2.4-3 48 DRR 10-3740 12/28/10 B 3.2.4-4 0 Amend. No. 123 12/18/99 B 3.2.4-5 48 DRR 10-3740 12/28/10 B 3.2.4-6 0 Amend. No. 123 12/18/99 B 3.2.4-7 48 DRR 10-3740 12/28/10 TAB - B 3.3 INSTRUMENTATION B 3.3.1-1 84 DRR 20-0400 08/18/20 B 3.3.1-2 0 Amend. No. 123 12/18/99 B 3.3.1-3 0 Amend. No. 123 12/18/99 B 3.3.1-4 0 Amend. No. 123 12/18/99 B 3.3.1-5 0 Amend. No. 123 12/18/99 B 3.3.1-6 0 Amend. No. 123 12/18/99 B 3.3.1-7 5 DRR 00-1427 10/12/00 B 3.3.1-8 0 Amend. No. 123 12/18/99 B 3.3.1-9 84 DRR 20-0400 08/18/20 B 3.3.1-10 84 DRR 20-0400 08/18/20 B 3.3.1-11 84 DRR 20-0400 08/18/20 B 3.3.1-12 84 DRR 20-0400 08/18/20 B 3.3.1-13 84 DRR 20-0400 08/18/20 B 3.3.1-14 84 DRR 20-0400 08/18/20 B 3.3.1-15 84 DRR 20-0400 08/18/20 B 3.3.1-16 84 DRR 20-0400 08/18/20 B 3.3.1-17 84 DRR 20-0400 08/18/20 B 3.3.1-18 84 DRR 20-0400 08/18/20 B 3.3.1-19 84 DRR 20-0400 08/18/20 B 3.3.1-20 84 DRR 20-0400 08/18/20 B 3.3.1-21 84 DRR 20-0400 08/18/20 B 3.3.1-22 84 DRR 20-0400 08/18/20 B 3.3.1-23 84 DRR 20-0400 08/18/20 B 3.3.1-24 84 DRR 20-0400 08/18/20 B 3.3.1-25 84 DRR 20-0400 08/18/20 B 3.3.1-26 84 DRR 20-0400 08/18/20 B 3.3.1-27 84 DRR 20-0400 08/18/20 B 3.3.1-28 84 DRR 20-0400 08/18/20 B 3.3.1-29 84 DRR 20-0400 08/18/20 B 3.3.1-30 84 DRR 20-0400 08/18/20 B 3.3.1-31 84 DRR 20-0400 08/18/20 B 3.3.1-32 84 DRR 20-0400 08/18/20 B 3.3.1-33 84 DRR 20-0400 08/18/20 Wolf Creek - Unit 1 iii Revision 87

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.1-34 84 DRR 20-0400 08/18/20 B 3.3.1-35 84 DRR 20-0400 08/18/20 B 3.3.1-36 84 DRR 20-0400 08/18/20 B 3.3.1-37 84 DRR 20-0400 08/18/20 B 3.3.1-38 84 DRR 20-0400 08/18/20 B 3.3.1-39 84 DRR 20-0400 08/18/20 B 3.3.1-40 84 DRR 20-0400 08/18/20 B 3.3.1-41 84 DRR 20-0400 08/18/20 B 3.3.1-42 84 DRR 20-0400 08/18/20 B 3.3.1-43 84 DRR 20-0400 08/18/20 B 3.3.1-44 84 DRR 20-0400 08/18/20 B 3.3.1-45 84 DRR 20-0400 08/18/20 B 3.3.1-46 84 DRR 20-0400 08/18/20 B 3.3.1-47 84 DRR 20-0400 08/18/20 B 3.3.1-48 84 DRR 20-0400 08/18/20 B 3.3.1-49 84 DRR 20-0400 08/18/20 B 3.3.1-50 84 DRR 20-0400 08/18/20 B 3.3.1-51 84 DRR 20-0400 08/18/20 B 3.3.1-52 84 DRR 20-0400 08/18/20 B 3.3.1-53 84 DRR 20-0400 08/18/20 B 3.3.1-54 84 DRR 20-0400 08/18/20 B 3.3.1-55 84 DRR 20-0400 08/18/20 B 3.3.1-56 84 DRR 20-0400 08/18/20 B 3.3.1-57 84 DRR 20-0400 08/18/20 B 3.3.1-58 84 DRR 20-0400 08/18/20 B 3.3.1-59 84 DRR 20-0400 08/18/20 B 3.3.1-60 84 DRR 20-0400 08/18/20 B 3.3.1-61 84 DRR 20-0400 08/18/20 B 3.3.1-62 84 DRR 20-0400 08/18/20 B 3.3.2-1 84 DRR 20-0400 08/18/20 B 3.3.2-2 0 Amend. No. 123 12/18/99 B 3.3.2-3 0 Amend. No. 123 12/18/99 B 3.3.2-4 0 Amend. No. 123 12/18/99 B 3.3.2-5 0 Amend. No. 123 12/18/99 B 3.3.2-6 7 DRR 01-0474 5/1/01 B 3.3.2-7 0 Amend. No. 123 12/18/99 B 3.3.2-8 0 Amend. No. 123 12/18/99 B 3.3.2-9 0 Amend. No. 123 12/18/99 B 3.3.2-10 0 Amend. No. 123 12/18/99 B 3.3.2-11 0 Amend. No. 123 12/18/99 B 3.3.2-12 81 DRR 19-1027 10/28/19 B 3.3.2-13 0 Amend. No. 123 12/18/99 B 3.3.2-14 2 DRR 00-0147 4/24/00 B 3.3.2-15 0 Amend. No. 123 12/18/99 B 3.3.2-16 0 Amend. No. 123 12/18/99 B 3.3.2-17 0 Amend. No. 123 12/18/99 B 3.3.2-18 0 Amend. No. 123 12/18/99 B 3.3.2-19 37 DRR 08-0503 4/8/08 B 3.3.2-20 37 DRR 08-0503 4/8/08 B 3.3.2-21 37 DRR 08-0503 4/8/08 B 3.3.2-22 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 iv Revision 87

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.2-23 37 DRR 08-0503 4/8/08 B 3.3.2-24 39 DRR 08-1096 8/28/08 B 3.3.2-25 39 DRR 08-1096 8/28/08 B 3.3.2-26 39 DRR 08-1096 8/28/08 B 3.3.2-27 37 DRR 08-0503 4/8/08 B 3.3.2-28 84 DRR 20-0400 08/18/20 B 3.3.2-29 0 Amend. No. 123 12/18/99 B 3.3.2-30 0 Amend. No. 123 12/18/99 B 3.3.2-31 52 DRR 11-0724 4/11/11 B 3.3.2-32 52 DRR 11-0724 4/11/11 B 3.3.2-33 0 Amend. No. 123 12/18/99 B 3.3.2-34 0 Amend. No. 123 12/18/99 B 3.3.2-35 20 DRR 04-1533 2/16/05 B 3.3.2-36 20 DRR 04-1533 2/16/05 B 3.3.2-37 20 DRR 04-1533 2/16/05 B 3.3.2-38 20 DRR 04-1533 2/16/05 B 3.3.2-39 25 DRR 06-0800 5/18/06 B 3.3.2-40 20 DRR 04-1533 2/16/05 B 3.3.2-41 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-42 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-43 20 DRR 04-1533 2/16/05 B 3.3.2-44 20 DRR 04-1533 2/16/05 B 3.3.2-45 20 DRR 04-1533 2/16/05 B 3.3.2-46 54 DRR 11-2394 11/16/11 B 3.3.2-47 43 DRR 09-1416 9/2/09 B 3.3.2-48 37 DRR 08-0503 4/8/08 B 3.3.2-49 20 DRR 04-1533 2/16/05 B 3.3.2-50 20 DRR 04-1533 2/16/05 B 3.3.2-51 43 DRR 09-1416 9/2/09 B 3.3.2-52 43 DRR 09-1416 9/2/09 B 3.3.2-53 43 DRR 09-1416 9/2/09 B 3.3.2-54 43 DRR 09-1416 9/2/09 B 3.3.2-55 43 DRR 09-1416 9/2/09 B 3.3.2-56 43 DRR 09-1416 9/2/09 B 3.3.2-57 43 DRR 09-1416 9/2/09 B 3.3.3-1 0 Amend. No. 123 12/18/99 B 3.3.3-2 5 DRR 00-1427 10/12/00 B 3.3.3-3 0 Amend. No. 123 12/18/99 B 3.3.3-4 0 Amend. No. 123 12/18/99 B 3.3.3-5 0 Amend. No. 123 12/18/99 B 3.3.3-6 8 DRR 01-1235 9/19/01 B 3.3.3-7 21 DRR 05-0707 4/20/05 B 3.3.3-8 81 DRR 19-1027 10/28/19 B 3.3.3-9 8 DRR 01-1235 9/19/01 B 3.3.3-10 19 DRR 04-1414 10/12/04 B 3.3.3-11 19 DRR 04-1414 10/12/04 B 3.3.3-12 21 DRR 05-0707 4/20/05 B 3.3.3-13 21 DRR 05-0707 4/20/05 B 3.3.3-14 8 DRR 01-1235 9/19/01 B 3.3.3-15 8 DRR 01-1235 9/19/01 B 3.3.4-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 v Revision 87

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.4-2 9 DRR 02-1023 2/28/02 B 3.3.4-3 15 DRR 03-0860 7/10/03 B 3.3.4-4 19 DRR 04-1414 10/12/04 B 3.3.4-5 1 DRR 99-1624 12/18/99 B 3.3.4-6 9 DRR 02-0123 2/28/02 B 3.3.5-1 0 Amend. No. 123 12/18/99 B 3.3.5-2 1 DRR 99-1624 12/18/99 B 3.3.5-3 1 DRR 99-1624 12/18/99 B 3.3.5-4 1 DRR 99-1624 12/18/99 B 3.3.5-5 0 Amend. No. 123 12/18/99 B 3.3.5-6 22 DRR 05-1375 6/28/05 B 3.3.5-7 22 DRR 05-1375 6/28/05 B 3.3.6-1 81 DRR 19-1027 10/28/19 B 3.3.6-2 81 DRR 19-1027 10/28/19 B 3.3.6-3 0 Amend. No. 123 12/18/99 B 3.3.6-4 0 Amend. No. 123 12/18/99 B 3.3.6-5 0 Amend. No. 123 12/18/99 B 3.3.6-6 0 Amend. No. 123 12/18/99 B 3.3.6-7 81 DRR 19-1027 10/28/19 B 3.3.7-1 81 DRR 19-1027 10/28/19 B 3.3.7-2 81 DRR 19-1027 10/28/19 B 3.3.7-3 57 DRR 13-0006 1/16/13 B 3.3.7-4 0 Amend. No. 123 12/18/99 B 3.3.7-5 0 Amend. No. 123 12/18/99 B 3.3.7-6 57 DRR 13-0006 1/16/13 B 3.3.7-7 0 Amend. No. 123 12/18/99 B 3.3.7-8 81 DRR 19-1027 10/28/19 B 3.3.8-1 84 DRR 20-0400 8/18/20 B 3.3.8-2 0 Amend. No. 123 12/18/99 B 3.3.8-3 57 DRR 13-0006 1/16/13 B 3.3.8-4 57 DRR 13-0006 1/16/13 B 3.3.8-5 0 Amend. No. 123 12/18/99 B 3.3.8-6 24 DRR 06-0051 2/28/06 B 3.3.8-7 81 DRR 19-1027 10/28/19 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1-1 84 DRR 20-0400 08/18/20 B 3.4.1-2 84 DRR 20-0400 08/18/20 B 3.4.1-3 10 DRR 02-0411 4/5/02 B 3.4.1-4 0 Amend. No. 123 12/18/99 B 3.4.1-5 0 Amend. No. 123 12/18/99 B 3.4.1-6 84 DRR 20-0400 08/18/20 B 3.4.2-1 0 Amend. No. 123 12/18/99 B 3.4.2-2 0 Amend. No. 123 12/18/99 B 3.4.2-3 0 Amend. No. 123 12/18/99 B 3.4.3-1 67 DRR 15-0116 2/10/15 B 3.4.3-2 0 Amend. No. 123 12/18/99 B 3.4.3-3 0 Amend. No. 123 12/18/99 B 3.4.3-4 0 Amend. No. 123 12/18/99 B 3.4.3-5 0 Amend. No. 123 12/18/99 B 3.4.3-6 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 vi Revision 87

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.3-7 0 Amend. No. 123 12/18/99 B 3.4.4-1 0 Amend. No. 123 12/18/99 B 3.4.4-2 29 DRR 06-1984 10/17/06 B 3.4.4-3 0 Amend. No. 123 12/18/99 B 3.4.5-1 0 Amend. No. 123 12/18/99 B 3.4.5-2 53 DRR 11-1513 7/18/11 B 3.4.5-3 29 DRR 06-1984 10/17/06 B 3.4.5-4 0 Amend. No. 123 12/18/99 B 3.4.5-5 12 DRR 02-1062 9/26/02 B 3.4.5-6 12 DRR 02-1062 9/26/02 B 3.4.6-1 53 DRR 11-1513 7/18/11 B 3.4.6-2 72 DRR 15-1918 10/26/15 B 3.4.6-3 12 DRR 02-1062 9/26/02 B 3.4.6-4 72 DRR 15-1918 10/26/15 B 3.4.6-5 75 DRR 16-1909 10/26/16 B 3.4.6-6 75 DRR 16-1909 10/26/16 B 3.4.7-1 12 DRR 02-1062 9/26/02 B 3.4.7-2 17 DRR 04-0453 5/26/04 B 3.4.7-3 72 DRR 15-1918 10/26/15 B 3.4.7-4 42 DRR 09-1009 7/16/09 B 3.4.7-5 72 DRR 15-1918 10/26/15 B 3.4.7-6 75 DRR 16-1909 10/26/16 B 3.4.8-1 53 DRR 11-1513 7/18/11 B 3.4.8-2 72 DRR 15-1918 10/26/15 B 3.4.8-3 42 DRR 09-1009 7/16/09 B 3.4.8-4 75 DRR 16-1909 10/26/16 B 3.4.8-5 72 DRR 15-1918 10/26/15 B 3.4.9-1 0 Amend. No. 123 12/18/99 B 3.4.9-2 0 Amend. No. 123 12/18/99 B 3.4.9-3 0 Amend. No. 123 12/18/99 B 3.4.9-4 0 Amend. No. 123 12/18/99 B 3.4.10-1 5 DRR 00-1427 10/12/00 B 3.4.10-2 5 DRR 00-1427 10/12/00 B 3.4.10-3 0 Amend. No. 123 12/18/99 B 3.4.10-4 32 DRR 07-0139 2/7/07 B 3.4.11-1 0 Amend. No. 123 12/18/99 B 3.4.11-2 1 DRR 99-1624 12/18/99 B 3.4.11-3 19 DRR 04-1414 10/12/04 B 3.4.11-4 0 Amend. No. 123 12/18/99 B 3.4.11-5 1 DRR 99-1624 12/18/99 B 3.4.11-6 0 Amend. No. 123 12/18/99 B 3.4.11-7 32 DRR 07-0139 2/7/07 B 3.4.12-1 61 DRR 14-0346 2/27/14 B 3.4.12-2 61 DRR 14-0346 2/27/14 B 3.4.12-3 0 Amend. No. 123 12/18/99 B 3.4.12-4 61 DRR 14-0346 2/27/14 B 3.4.12-5 61 DRR 14-0346 2/27/14 B 3.4.12-6 56 DRR 12-1792 11/7/12 B 3.4.12-7 61 DRR 14-0346 2/27/14 B 3.4.12-8 1 DRR 99-1624 12/18/99 B 3.4.12-9 56 DRR 12-1792 11/7/12 Wolf Creek - Unit 1 vii Revision 87

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.12-10 0 Amend. No. 123 12/18/99 B 3.4.12-11 61 DRR 14-0346 2/27/14 B 3.4.12-12 32 DRR 07-0139 2/7/07 B 3.4.12-13 0 Amend. No. 123 12/18/99 B 3.4.12-14 32 DRR 07-0139 2/7/07 B 3.4.13-1 0 Amend. No. 123 12/18/99 B 3.4.13-2 81 DRR 19-1027 10/28/19 B 3.4.13-3 29 DRR 06-1984 10/17/06 B 3.4.13-4 35 DRR 07-1553 9/28/07 B 3.4.13-5 35 DRR 07-1553 9/28/07 B 3.4.13-6 81 DRR 19-1027 10/28/19 B 3.4.14-1 0 Amend. No. 123 12/18/99 B 3.4.14-2 0 Amend. No. 123 12/18/99 B 3.4.14-3 0 Amend. No. 123 12/18/99 B 3.4.14-4 0 Amend. No. 123 12/18/99 B 3.4.14-5 32 DRR 07-0139 2/7/07 B 3.4.14-6 32 DRR 07-0139 2/7/07 B 3.4.15-1 31 DRR 06-2494 12/13/06 B 3.4.15-2 31 DRR 06-2494 12/13/06 B 3.4.15-3 33 DRR 07-0656 5/1/07 B 3.4.15-4 33 DRR 07-0656 5/1/07 B 3.4.15-5 65 DRR 14-2146 9/30/14 B 3.4.15-6 31 DRR 06-2494 12/13/06 B 3.4.15-7 31 DRR 06-2494 12/13/06 B 3.4.15-8 31 DRR 06-2494 12/13/06 B 3.4.16-1 81 DRR 19-1027 10/28/19 B 3.4.16-2 84 DRR 20-0400 08/18/20 B 3.4.16-3 31 DRR 06-2494 12/13/06 B 3.4.16-4 31 DRR 06-2494 12/13/06 B 3.4.16-5 81 DRR 19-1027 10/28/19 B 3.4.17-1 29 DRR 06-1984 10/17/06 B 3.4.17-2 81 DRR 19-1027 10/28/19 B 3.4.17-3 52 DRR 11-0724 4/11/11 B 3.4.17-4 81 DRR 19-1027 10/28/19 B 3.4.17-5 57 DRR 13-0006 1/16/13 B 3.4.17-6 57 DRR 13-0006 1/16/13 B 3.4.17-7 81 DRR 19-1027 10/28/19 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1-1 0 Amend. No. 123 12/18/99 B 3.5.1-2 0 Amend. No. 123 12/18/99 B 3.5.1-3 73 DRR 15-2135 11/17/15 B 3.5.1-4 73 DRR 15-2135 11/17/15 B 3.5.1-5 1 DRR 99-1624 12/18/99 B 3.5.1-6 1 DRR 99-1624 12/18/99 B 3.5.1-7 71 DRR 15-1528 7/30/15 B 3.5.1-8 1 DRR 99-1624 12/18/99 B 3.5.2-1 84 DRR 20-0400 08/18/20 B 3.5.2-2 0 Amend. No. 123 12/18/99 B 3.5.2-3 0 Amend. No. 123 12/18/99 B 3.5.2-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 viii Revision 87

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TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

B 3.5.2-5 72 DRR 15-1918 10/26/15 B 3.5.2-6 42 DRR 09-1009 7/16/09 B 3.5.2-7 42 DRR 09-1009 7/16/09 B 3.5.2-8 72 DRR 15-1918 10/26/15 B 3.5.2-9 75 DRR 16-1909 10/26/16 B 3.5.2-10 80 DRR 19-0524 5/30/19 B 3.5.2-11 72 DRR 15-1918 10/26/15 B 3.5.2-12 72 DRR 15-1918 10/26/15 B 3.5.3-1 56 DRR 12-1792 11/7/12 B 3.5.3-2 72 DRR 15-1918 10/26/15 B 3.5.3-3 56 DRR 12-1792 11/7/12 B 3.5.3-4 56 DRR 12-1792 11/7/12 B 3.5.4-1 0 Amend. No. 123 12/18/99 B 3.5.4-2 0 Amend. No. 123 12/18/99 B 3.5.4-3 0 Amend. No. 123 12/18/99 B 3.5.4-4 0 Amend. No. 123 12/18/99 B 3.5.4-5 0 Amend. No. 123 12/18/99 B 3.5.4-6 26 DRR 06-1350 7/24/06 B 3.5.5-1 21 DRR 05-0707 4/20/05 B 3.5.5-2 21 DRR 05-0707 4/20/05 B 3.5.5-3 2 Amend. No. 132 4/24/00 B 3.5.5-4 21 DRR 05-0707 4/20/05 TAB - B 3.6 CONTAINMENT SYSTEMS B 3.6.1-1 0 Amend. No. 123 12/18/99 B 3.6.1-2 81 DRR 19-1027 10/28/19 B 3.6.1-3 0 Amend. No. 123 12/18/99 B 3.6.1-4 87 DRR 21-0359 3/25/21 B 3.6.2-1 81 DRR 19-1027 10/28/19 B 3.6.2-2 0 Amend. No. 123 12/18/99 B 3.6.2-3 0 Amend. No. 123 12/18/99 B 3.6.2-4 0 Amend. No. 123 12/18/99 B 3.6.2-5 0 Amend. No. 123 12/18/99 B 3.6.2-6 0 Amend. No. 123 12/18/99 B 3.6.2-7 0 Amend. No. 123 12/18/99 B 3.6.3-1 0 Amend. No. 123 12/18/99 B 3.6.3-2 84 DRR 20-0400 08/18/20 B 3.6.3-3 81 DRR 19-1027 10/28/19 B 3.6.3-4 49 DRR 11-0014 1/31/11 B 3.6.3-5 49 DRR 11-0014 1/31/11 B 3.6.3-6 49 DRR 11-0014 1/31/11 B 3.6.3-7 41 DRR 09-0288 3/20/09 B 3.6.3-8 36 DRR 08-0255 3/11/08 B 3.6.3-9 36 DRR 08-0255 3/11/08 B 3.6.3-10 8 DRR 01-1235 9/19/01 B 3.6.3-11 36 DRR 08-0255 3/11/08 B 3.6.3-12 36 DRR 08-0255 3/11/08 B 3.6.3-13 50 DRR 11-0449 3/9/11 B 3.6.3-14 36 DRR 08-0255 3/11/08 B 3.6.3-15 39 DRR 08-1096 8/28/08 B 3.6.3-16 39 DRR 08-1096 8/28/08 Wolf Creek - Unit 1 ix Revision 87

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TAB - B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.3-17 36 DRR 08-0255 3/11/08 B 3.6.3-18 36 DRR 08-0255 3/11/08 B 3.6.3-19 36 DRR 08-0255 3/11/08 B 3.6.4-1 39 DRR 08-1096 8/28/08 B 3.6.4-2 0 Amend. No. 123 12/18/99 B 3.6.4-3 0 Amend. No. 123 12/18/99 B 3.6.5-1 0 Amend. No. 123 12/18/99 B 3.6.5-2 37 DRR 08-0503 4/8/08 B 3.6.5-3 13 DRR 02-1458 12/03/02 B 3.6.5-4 0 Amend. No. 123 12/18/99 B 3.6.6-1 81 DRR 19-1027 10/28/19 B 3.6.6-2 63 DRR 14-1572 7/1/14 B 3.6.6-3 37 DRR 08-0503 4/8/08 B 3.6.6-4 81 DRR 19-1027 10/28/19 B 3.6.6-5 0 Amend. No. 123 12/18/99 B 3.6.6-6 18 DRR 04-1018 9/1/04 B 3.6.6-7 72 DRR 15-1918 10/26/15 B 3.6.6-8 80 DRR 19-0524 5/30/19 B 3.6.6-9 72 DRR 15-1918 10/26/15 B 3.6.6-10 75 DRR 16-1909 10/26/16 B 3.6.6.11 80 DRR 19-0524 5/30/19 B 3.6.7-1 0 Amend. No. 123 12/18/99 B 3.6.7-2 81 DRR 19-1027 10/28/19 B 3.6.7-3 81 DRR 19-1027 10/28/19 B 3.6.7-4 81 DRR 19-1027 10/28/19 B 3.6.7-5 42 DRR 09-1009 7/16/09 TAB - B 3.7 PLANT SYSTEMS B 3.7.1-1 0 Amend. No. 123 12/18/99 B 3.7.1-2 84 DRR 20-0400 08/18/20 B 3.7.1-3 0 Amend. No. 123 12/18/99 B 3.7.1-4 84 DRR 20-0400 08/18/20 B 3.7.1-5 84 DRR 20-0400 08/18/20 B 3.7.1-6 84 DRR 20-0400 08/18/20 B 3.7.2-1 44 DRR 09-1744 10/28/09 B 3.7.2-2 44 DRR 09-1744 10/28/09 B 3.7.2-3 44 DRR 09-1744 10/28/09 B 3.7.2-4 81 DRR 19-1027 10/28/19 B 3.7.2-5 44 DRR 09-1744 10/28/09 B 3.7.2-6 44 DRR 09-1744 10/28/09 B 3.7.2-7 44 DRR 09-1744 10/28/09 B 3.7.2-8 44 DRR 09-1744 10/28/09 B 3.7.2-9 44 DRR 09-1744 10/28/09 B 3.7.2-10 81 DRR 19-1027 10/28/19 B 3.7.2-11 44 DRR 09-1744 10/28/09 B 3.7.3-1 37 DRR 08-0503 4/8/08 B 3.7.3-2 50 DRR 11-0449 3/9/11 B 3.7.3-3 37 DRR 08-0503 4/8/08 B 3.7.3-4 37 DRR 08-0503 4/8/08 B 3.7.3-5 37 DRR 08-0503 4/8/08 B 3.7.3-6 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 x Revision 87

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.3-7 37 DRR 08-0503 4/8/08 B 3.7.3-8 37 DRR 08-0503 4/8/08 B 3.7.3-9 66 DRR 14-2329 11/6/14 B 3.7.3-10 66 DRR 14-2329 11/6/14 B 3.7.3-11 37 DRR 08-0503 4/8/08 B 3.7.4-1 1 DRR 99-1624 12/18/99 B 3.7.4-2 84 DRR 20-0400 08/18/20 B 3.7.4-3 19 DRR 04-1414 10/12/04 B 3.7.4-4 19 DRR 04-1414 10/12/04 B 3.7.4-5 84 DRR 20-0400 08/18/20 B 3.7.5-1 54 DRR 11-2394 11/16/11 B 3.7.5-2 54 DRR 11-2394 11/16/11 B 3.7.5-3 0 Amend. No. 123 12/18/99 B 3.7.5-4 85 DRR 20-0988 10/24/20 B 3.7.5-5 76 DRR 17-0343 2/21/17 B 3.7.5-6 85 DRR 20-0988 10/24/20 B 3.7.5-7 85 DRR 20-0988 10/24/20 B 3.7.5-8 85 DRR 20-0988 10/24/20 B 3.7.5-9 85 DRR 20-0988 10/24/20 B 3.7.5-10 85 DRR 20-0988 10/24/20 B 3.7.6-1 0 Amend. No. 123 12/18/99 B 3.7.6-2 0 Amend. No. 123 12/18/99 B 3.7.6-3 0 Amend. No. 123 12/18/99 B 3.7.7-1 0 Amend. No. 123 12/18/99 B 3.7.7-2 0 Amend. No. 123 12/18/99 B 3.7.7-3 0 Amend. No. 123 12/18/99 B 3.7.7-4 1 DRR 99-1624 12/18/99 B 3.7.8-1 0 Amend. No. 123 12/18/99 B 3.7.8-2 0 Amend. No. 123 12/18/99 B 3.7.8-3 0 Amend. No. 123 12/18/99 B 3.7.8-4 0 Amend. No. 123 12/18/99 B 3.7.8-5 0 Amend. No. 123 12/18/99 B 3.7.9-1 3 Amend. No. 134 7/14/00 B 3.7.9-2 3 Amend. No. 134 7/14/00 B 3.7.9-3 3 Amend. No. 134 7/14/00 B 3.7.9-4 3 Amend. No. 134 7/14/00 B 3.7.10-1 64 DRR 14-1822 8/28/14 B 3.7.10-2 81 DRR 19-1027 10/28/19 B 3.7.10-3 81 DRR 19-1027 10/28/19 B 3.7.10-4 81 DRR 19-1027 10/28/19 B 3.7.10-5 81 DRR 19-1027 10/28/19 B 3.7.10-6 57 DRR 13-0006 1/16/13 B 3.7.10-7 64 DRR 14-1822 8/28/14 B 3.7.10-8 81 DRR 19-1027 10/28/19 B 3.7.10-9 81 DRR 19-1027 10/28/19 B 3.7.11-1 0 Amend. No. 123 12/18/99 B 3.7.11-2 57 DRR 13-0006 1/16/13 B 3.7.11-3 63 DRR 14-1572 7/1/14 B 3.7.11-4 63 DRR 14-1572 7/1/14 B 3.7.12-1 0 Amend. No. 123 12/18/99 B 3.7.13-1 24 DRR 06-0051 2/28/06 Wolf Creek - Unit 1 xi Revision 87

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.13-2 81 DRR 19-1027 10/28/19 B 3.7.13-3 81 DRR 19-1027 10/28/19 B 3.7.13-4 81 DRR 19-1027 10/28/19 B 3.7.13-5 81 DRR 19-1027 10/28/19 B 3.7.13-6 81 DRR 19-1027 10/28/19 B 3.7.13-7 81 DRR 19-1027 10/28/19 B 3.7.13-8 81 DRR 19-1027 10/28/19 B 3.7.14-1 0 Amend. No. 123 12/18/99 B 3.7.15-1 81 DRR 19-1027 10/28/19 B 3.7.15-2 81 DRR 19-1027 10/28/19 B 3.7.15-3 81 DRR 19-1027 10/28/19 B 3.7.16-1 5 DRR 00-1427 10/12/00 B 3.7.16-2 23 DRR 05-1995 9/28/05 B 3.7.16-3 5 DRR 00-1427 10/12/00 B 3.7.17-1 7 DRR 01-0474 5/1/01 B 3.7.17-2 7 DRR 01-0474 5/1/01 B 3.7.17-3 5 DRR 00-1427 10/12/00 B 3.7.18-1 81 DRR 19-1027 10/28/19 B 3.7.18-2 81 DRR 19-1027 10/28/19 B 3.7.18-3 81 DRR 19-1027 10/28/19 B 3.7.19-1 44 DRR 09-1744 10/28/09 B 3.7.19-2 54 DRR 11-2394 11/16/11 B 3.7.19-3 54 DRR 11-2394 11/16/11 B 3.7.19-4 61 DRR 14-0346 2/27/14 B 3.7.19-5 61 DRR 14-0346 2/27/14 B 3.7.19-6 54 DRR 11-2394 11/16/11 B 3.7.19-7 54 DRR 11-2394 11/16/11 B 3.7.20-1 79 DRR 18-1579 10/22/18 B 3.7.20-2 79 DRR 18-1579 10/22/18 B 3.7.20-3 85 DRR 20-0988 10/24/20 B 3.7.20-4 79 DRR 18-1579 10/22/18 B 3.7.20-5 79 DRR 18-1579 10/22/18 TAB - B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1-1 54 DRR 11-2394 11/16/11 B 3.8.1-2 0 Amend. No. 123 12/18/99 B 3.8.1-3 86 DRR 21-0212 2/18/21 B 3.8.1-4 86 DRR 21-0212 2/18/21 B 3.8.1-5 59 DRR 13-1524 6/26/13 B 3.8.1-6 25 DRR 06-0800 5/18/06 B 3.8.1-7 26 DRR 06-1350 7/24/06 B 3.8.1-8 35 DRR 07-1553 9/28/07 B 3.8.1-9 42 DRR 09-1009 7/16/09 B 3.8.1-10 39 DRR 08-1096 8/28/08 B 3.8.1-11 36 DRR 08-0255 3/11/08 B 3.8.1-12 75 DRR 16-1909 10/26/16 B 3.8.1-13 47 DRR 10-1089 6/16/10 B 3.8.1-14 47 DRR 10-1089 6/16/10 B 3.8.1-15 47 DRR 10-1089 6/16/10 B 3.8.1-16 26 DRR 06-1350 7/24/06 B 3.8.1-17 26 DRR 06-1350 7/24/06 Wolf Creek - Unit 1 xii Revision 87

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IMPLEMENTED (4)

TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.1-18 59 DRR 13-1524 6/26/13 B 3.8.1-19 26 DRR 06-1350 7/24/06 B 3.8.1-20 26 DRR 06-1350 7/24/06 B 3.8.1-21 33 DRR 07-0656 5/1/07 B 3.8.1-22 33 DRR 07-0656 5/1/07 B 3.8.1-23 74 DRR 16-1182 7/7/16 B 3.8.1-24 74 DRR 16-1182 7/7/16 B 3.8.1-25 74 DRR 16-1182 7/7/16 B 3.8.1-26 74 DRR 16-1182 7/7/16 B 3.8.1-27 74 DRR 16-1182 7/7/16 B 3.8.1-28 74 DRR 16-1182 7/7/16 B 3.8.1-29 74 DRR 16-1182 7/7/16 B 3.8.1-30 74 DRR 16-1182 7/7/16 B 3.8.1-31 74 DRR 16-1182 7/7/16 B 3.8.1-32 74 DRR 16-1182 7/7/16 B 3.8.1-33 74 DRR 16-1182 7/7/16 B 3.8.1-34 74 DRR 16-1182 7/7/16 B 3.8.2-1 57 DRR 13-0006 1/16/13 B 3.8.2-2 0 Amend. No. 123 12/18/99 B 3.8.2-3 86 DRR 21-0212 2/18/21 B 3.8.2-4 86 DRR 21-0212 2/18/21 B 3.8.2-5 57 DRR 13-0006 1/16/13 B 3.8.2-6 57 DRR 13-0006 1/16/13 B 3.8.2-7 57 DRR 13-0006 1/16/13 B 3.8.3-1 1 DRR 99-1624 12/18/99 B 3.8.3-2 0 Amend. No. 123 12/18/99 B 3.8.3-3 0 Amend. No. 123 12/18/99 B 3.8.3-4 1 DRR 99-1624 12/18/99 B 3.8.3-5 0 Amend. No. 123 12/18/99 B 3.8.3-6 0 Amend. No. 123 12/18/99 B 3.8.3-7 12 DRR 02-1062 9/26/02 B 3.8.3-8 1 DRR 99-1624 12/18/99 B 3.8.3-9 0 Amend. No. 123 12/18/99 B 3.8.4-1 0 Amend. No. 123 12/18/99 B 3.8.4-2 0 Amend. No. 123 12/18/99 B 3.8.4-3 0 Amend. No. 123 12/18/99 B 3.8.4-4 0 Amend. No. 123 12/18/99 B 3.8.4-5 50 DRR 11-0449 3/9/11 B 3.8.4-6 50 DRR 11-0449 3/9/11 B 3.8.4-7 6 DRR 00-1541 3/13/01 B 3.8.4-8 0 Amend. No. 123 12/18/99 B 3.8.4-9 2 DRR 00-0147 4/24/00 B 3.8.5-1 57 DRR 13-0006 1/16/13 B 3.8.5-2 0 Amend. No. 123 12/18/99 B 3.8.5-3 57 DRR 13-0006 1/16/13 B 3.8.5-4 57 DRR 13-0006 1/16/13 B 3.8.5-5 57 DRR 13-0006 1/16/13 B 3.8.6-1 0 Amend. No. 123 12/18/99 B 3.8.6-2 0 Amend. No. 123 12/18/99 B 3.8.6-3 0 Amend. No. 123 12/18/99 B 3.8.6-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiii Revision 87

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.6-5 0 Amend. No. 123 12/18/99 B 3.8.6-6 0 Amend. No. 123 12/18/99 B 3.8.7-1 69 DRR 15-0493 3/26/15 B 3.8.7-2 69 DRR 15-0493 3/26/15 B 3.8.7-3 69 DRR 15-0493 3/26/15 B 3.8.7-4 0 Amend. No. 123 12/18/99 B 3.8.8-1 57 DRR 13-0006 1/16/13 B 3.8.8-2 0 Amend. No. 123 12/18/99 B 3.8.8-3 69 DRR 15-0493 3/26/15 B 3.8.8-4 57 DRR 13-0006 1/16/13 B 3.8.8-5 69 DRR 15-0493 3/26/15 B 3.8.9-1 54 DRR 11-2394 11/16/11 B 3.8.9-2 69 DRR 15-0493 3/26/15 B 3.8.9-3 54 DRR 11-2394 11/16/11 B 3.8.9-4 0 Amend. No. 123 12/18/99 B 3.8.9-5 69 DRR 15-0493 3/26/15 B 3.8.9-6 0 Amend. No. 123 12/18/99 B 3.8.9-7 0 Amend. No. 123 12/18/99 B 3.8.9-8 1 DRR 99-1624 12/18/99 B 3.8.9-9 0 Amend. No. 123 12/18/99 B 3.8.10-1 57 DRR 13-0006 1/16/13 B 3.8.10-2 0 Amend. No. 123 12/18/99 B 3.8.10-3 0 Amend. No. 123 12/18/99 B 3.8.10-4 57 DRR 13-0006 1/16/13 B 3.8.10-5 57 DRR 13-0006 1/16/13 B 3.8.10-6 57 DRR 13-0006 1/16/13 TAB - B 3.9 REFUELING OPERATIONS B 3.9.1-1 0 Amend. No. 123 12/18/99 B 3.9.1-2 19 DRR 04-1414 10/12/04 B 3.9.1-3 19 DRR 04-1414 10/12/04 B 3.9.1-4 19 DRR 04-1414 10/12/04 B 3.9.2-1 0 Amend. No. 123 12/18/99 B 3.9.2-2 0 Amend. No. 123 12/18/99 B 3.9.2-3 0 Amend. No. 123 12/18/99 B 3.9.3-1 68 DRR 15-0248 2/26/15 B 3.9.3-2 68 DRR 15-0248 2/26/15 B 3.9.3-3 51 DRR 11-0664 3/21/11 B 3.9.3-4 68 DRR 15-0248 2/26/15 B 3.9.4-1 81 DRR 19-1027 10/28/19 B 3.9.4-2 13 DRR 02-1458 12/03/02 B 3.9.4-3 81 DRR 19-1027 10/28/19 B 3.9.4-4 23 DRR 05-1995 9/28/05 B 3.9.4-5 33 DRR 07-0656 5/1/07 B 3.9.4-6 23 DRR 05-1995 9/28/05 B 3.9.5-1 0 Amend. No. 123 12/18/99 B 3.9.5-2 72 DRR 15-1918 10/26/15 B 3.9.5-3 32 DRR 07-0139 2/7/07 B 3.9.5-4 75 DRR 16-1909 10/26/16 B 3.9.5-5 75 DRR 16-1909 10/26/16 B 3.9.6-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiv Revision 87

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TAB - B 3.9 REFUELING OPERATIONS (continued)

B 3.9.6-2 72 DRR 15-1918 10/26/15 B 3.9.6-3 42 DRR 09-1009 7/16/09 B 3.9.6-4 72 DRR 15-1918 10/26/15 B 3.9.6-5 75 DRR 16-1909 10/26/16 B 3.9.7-1 81 DRR 19-1027 10/28/19 B 3.9.7-2 81 DRR 19-1027 10/28/19 B 3.9.7-3 81 DRR 19-1027 10/28/19 Note 1 The page number is listed on the center of the bottom of each page.

Note 2 The revision number is listed in the lower right hand corner of each page. The Revision number will be page specific.

Note 3 The change document will be the document requesting the change. Amendment No.

123 issued the improved Technical Specifications and associated Bases which affected each page. The NRC has indicated that Bases changes will not be issued with License Amendments. Therefore, the change document should be a DRR number in accordance with AP 26A-002.

Note 4 The date effective or implemented is the date the Bases pages are issued by Document Control.

Wolf Creek - Unit 1 xv Revision 87

ENCLOSURE Page 35 of 264 Containment B 3.6.1 BASES SURVEILLANCE SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. The containment concrete visual examinations may be performed during either power operation, e.g., performed concurrently with other containment inspection-related activities such as tendon testing, or during a maintenance/refueling outage. The visual examinations of the steel liner plate inside containment are performed during maintenance or refueling outages since this is the only time the liner plate is fully accessible.

Failure to meet air lock and purge valve with resilient seal leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A, B, and C leakage causes that to exceed limits. As left leakage prior to the first startup after performing a required Containment Leakage Rate Testing Program leakage test is required to be < 0.6 La for combined Type B and C leakage, and 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of 1.0 La. At 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. Prior to entering Mode 4, containment as-left leakage summation is required to be less than < 0.6 La for combined Type B and C leakage based on the maximum pathway leakage summation. This is specific to transitioning from Mode 5 to Mode 4.

SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.

SR 3.6.1.2 This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program. Testing and Frequency are in accordance with ASME Code Section XI, Subsection IWL (Ref. 4), and applicable addenda as required by 10 CFR 50.55a, except where an exemption or relief has been authorized by the NRC.

REFERENCES 1. 10 CFR 50, Appendix J, Option B.

2. USAR, Chapter 15.

3. USAR, Section 6.2.

4. ASME Code Section XI, Subsection IWL.

Wolf Creek - Unit 1 B 3.6.1-4 Revision 87

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TAB - Title Page Technical Specification Cover Page Title Page TAB - Table of Contents i 34 DRR 07-1057 7/10/07 ii 29 DRR 06-1984 10/17/06 iii 44 DRR 09-1744 10/28/09 TAB - B 2.0 SAFETY LIMITS (SLs)

B 2.1.1-1 0 Amend. No. 123 12/18/99 B 2.1.1-2 14 DRR 03-0102 2/12/03 B 2.1.1-3 14 DRR 03-0102 2/12/03 B 2.1.1-4 0 Amend. No. 123 2/12/03 B 2.1.2-1 84 DRR 20-0400 08/18/20 B 2.1.2-2 84 DRR 20-0400 08/18/20 B 2.1.2-3 81 DRR 19-1027 10/28/19 TAB - B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 81 DRR 19-1027 10/28/19 B 3.0-2 0 Amend. No. 123 12/18/99 B 3.0-3 81 DRR 19-1027 10/28/19 B 3.0-4 81 DRR 19-1027 10/28/19 B 3.0-5 81 DRR 19-1027 10/28/19 B 3.0-6 81 DRR 19-1027 10/28/19 B 3.0-7 81 DRR 19-1027 10/28/19 B 3.0-8 81 DRR 19-1027 10/28/19 B 3.0-9 81 DRR 19-1027 10/28/19 B 3.0-10 81 DRR 19-1027 10/28/19 B 3.0-11 81 DRR 19-1027 10/28/19 B 3.0-12 81 DRR 19-1027 10/28/19 B 3.0-13 81 DRR 19-1027 10/28/19 B 3.0-14 81 DRR 19-1027 10/28/19 B 3.0-15 81 DRR 19-1027 10/28/19 B 3.0-16 81 DRR 19-1027 10/28/19 B 3.0-17 81 DRR 19-1027 10/28/19 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1-1 0 Amend. No. 123 12/18/99 B 3.1.1-2 0 Amend. No. 123 12/18/99 B 3.1.1-3 0 Amend. No. 123 12/18/99 B 3.1.1-4 81 DRR 19-1027 10/28/19 B 3.1.1-5 81 DRR 19-1027 10/28/19 B 3.1.2-1 0 Amend. No. 123 12/18/99 B 3.1.2-2 0 Amend. No. 123 12/18/99 B 3.1.2-3 0 Amend. No. 123 12/18/99 B 3.1.2-4 0 Amend. No. 123 12/18/99 B 3.1.2-5 0 Amend. No. 123 12/18/99 B 3.1.3-1 0 Amend. No. 123 12/18/99 B 3.1.3-2 0 Amend. No. 123 12/18/99 B 3.1.3-3 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 i Revision 88

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TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)

B 3.1.3-4 0 Amend. No. 123 12/18/99 B 3.1.3-5 0 Amend. No. 123 12/18/99 B 3.1.3-6 0 Amend. No. 123 12/18/99 B 3.1.4-1 0 Amend. No. 123 12/18/99 B 3.1.4-2 0 Amend. No. 123 12/18/99 B 3.1.4-3 48 DRR 10-3740 12/28/10 B 3.1.4-4 0 Amend. No. 123 12/18/99 B 3.1.4-5 0 Amend. No. 123 12/18/99 B 3.1.4-6 48 DRR 10-3740 12/28/10 B 3.1.4-7 0 Amend. No. 123 12/18/99 B 3.1.4-8 0 Amend. No. 123 12/18/99 B 3.1.4-9 0 Amend. No. 123 12/18/99 B 3.1.5-1 0 Amend. No. 123 12/18/99 B 3.1.5-2 0 Amend. No. 123 12/18/99 B 3.1.5-3 0 Amend. No. 123 12/18/99 B 3.1.5-4 0 Amend. No. 123 12/18/99 B 3.1.6-1 0 Amend. No. 123 12/18/99 B 3.1.6-2 0 Amend. No. 123 12/18/99 B 3.1.6-3 0 Amend. No. 123 12/18/99 B 3.1.6-4 0 Amend. No. 123 12/18/99 B 3.1.6-5 0 Amend. No. 123 12/18/99 B 3.1.6-6 0 Amend. No. 123 12/18/99 B 3.1.7-1 0 Amend. No. 123 12/18/99 B 3.1.7-2 0 Amend. No. 123 12/18/99 B 3.1.7-3 48 DRR 10-3740 12/28/10 B 3.1.7-4 48 DRR 10-3740 12/28/10 B 3.1.7-5 48 DRR 10-3740 12/28/10 B 3.1.7-6 0 Amend. No. 123 12/18/99 B 3.1.8-1 0 Amend. No. 123 12/18/99 B 3.1.8-2 0 Amend. No. 123 12/18/99 B 3.1.8-3 15 DRR 03-0860 7/10/03 B 3.1.8-4 15 DRR 03-0860 7/10/03 B 3.1.8-5 0 Amend. No. 123 12/18/99 B 3.1.8-6 5 DRR 00-1427 10/12/00 B 3.1.9-1 84 DRR 20-0400 08/18/20 B 3.1.9-2 84 DRR 20-0400 08/18/20 B 3.1.9-3 84 DRR 20-0400 08/18/20 B 3.1.9-4 84 DRR 20-0400 08/18/20 B 3.1.9-5 84 DRR 20-0400 08/18/20 TAB - B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1-1 48 DRR 10-3740 12/28/10 B 3.2.1-2 0 Amend. No. 123 12/18/99 B 3.2.1-3 48 DRR 10-3740 12/28/10 B 3.2.1-4 48 DRR 10-3740 12/28/10 B 3.2.1-5 48 DRR 10-3740 12/28/10 B 3.2.1-6 48 DRR 10-3740 12/28/10 B 3.2.1-7 48 DRR 10-3740 12/28/10 B 3.2.1-8 48 DRR 10-3740 12/28/10 B 3.2.1-9 29 DRR 06-1984 10/17/06 B 3.2.1-10 70 DRR 15-0944 4/28/15 Wolf Creek - Unit 1 ii Revision 88

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TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued)

B 3.2.2-1 48 DRR 10-3740 12/28/10 B 3.2.2-2 0 Amend. No. 123 12/18/99 B 3.2.2-3 48 DRR 10-3740 12/28/10 B 3.2.2-4 48 DRR 10-3740 12/28/10 B 3.2.2-5 48 DRR 10-3740 12/28/10 B 3.2.2-6 70 DRR 15-0944 4/28/15 B 3.2.3-1 0 Amend. No. 123 12/18/99 B 3.2.3-2 0 Amend. No. 123 12/18/99 B 3.2.3-3 0 Amend. No. 123 12/18/99 B 3.2.4-1 0 Amend. No. 123 12/18/99 B 3.2.4-2 0 Amend. No. 123 12/18/99 B 3.2.4-3 48 DRR 10-3740 12/28/10 B 3.2.4-4 0 Amend. No. 123 12/18/99 B 3.2.4-5 48 DRR 10-3740 12/28/10 B 3.2.4-6 0 Amend. No. 123 12/18/99 B 3.2.4-7 48 DRR 10-3740 12/28/10 TAB - B 3.3 INSTRUMENTATION B 3.3.1-1 84 DRR 20-0400 08/18/20 B 3.3.1-2 0 Amend. No. 123 12/18/99 B 3.3.1-3 0 Amend. No. 123 12/18/99 B 3.3.1-4 0 Amend. No. 123 12/18/99 B 3.3.1-5 0 Amend. No. 123 12/18/99 B 3.3.1-6 0 Amend. No. 123 12/18/99 B 3.3.1-7 5 DRR 00-1427 10/12/00 B 3.3.1-8 0 Amend. No. 123 12/18/99 B 3.3.1-9 84 DRR 20-0400 08/18/20 B 3.3.1-10 84 DRR 20-0400 08/18/20 B 3.3.1-11 84 DRR 20-0400 08/18/20 B 3.3.1-12 84 DRR 20-0400 08/18/20 B 3.3.1-13 84 DRR 20-0400 08/18/20 B 3.3.1-14 84 DRR 20-0400 08/18/20 B 3.3.1-15 84 DRR 20-0400 08/18/20 B 3.3.1-16 84 DRR 20-0400 08/18/20 B 3.3.1-17 84 DRR 20-0400 08/18/20 B 3.3.1-18 84 DRR 20-0400 08/18/20 B 3.3.1-19 84 DRR 20-0400 08/18/20 B 3.3.1-20 84 DRR 20-0400 08/18/20 B 3.3.1-21 84 DRR 20-0400 08/18/20 B 3.3.1-22 84 DRR 20-0400 08/18/20 B 3.3.1-23 84 DRR 20-0400 08/18/20 B 3.3.1-24 84 DRR 20-0400 08/18/20 B 3.3.1-25 84 DRR 20-0400 08/18/20 B 3.3.1-26 84 DRR 20-0400 08/18/20 B 3.3.1-27 84 DRR 20-0400 08/18/20 B 3.3.1-28 84 DRR 20-0400 08/18/20 B 3.3.1-29 84 DRR 20-0400 08/18/20 B 3.3.1-30 84 DRR 20-0400 08/18/20 B 3.3.1-31 84 DRR 20-0400 08/18/20 B 3.3.1-32 84 DRR 20-0400 08/18/20 B 3.3.1-33 84 DRR 20-0400 08/18/20 Wolf Creek - Unit 1 iii Revision 88

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.1-34 84 DRR 20-0400 08/18/20 B 3.3.1-35 84 DRR 20-0400 08/18/20 B 3.3.1-36 84 DRR 20-0400 08/18/20 B 3.3.1-37 84 DRR 20-0400 08/18/20 B 3.3.1-38 84 DRR 20-0400 08/18/20 B 3.3.1-39 84 DRR 20-0400 08/18/20 B 3.3.1-40 84 DRR 20-0400 08/18/20 B 3.3.1-41 84 DRR 20-0400 08/18/20 B 3.3.1-42 84 DRR 20-0400 08/18/20 B 3.3.1-43 84 DRR 20-0400 08/18/20 B 3.3.1-44 84 DRR 20-0400 08/18/20 B 3.3.1-45 84 DRR 20-0400 08/18/20 B 3.3.1-46 84 DRR 20-0400 08/18/20 B 3.3.1-47 84 DRR 20-0400 08/18/20 B 3.3.1-48 84 DRR 20-0400 08/18/20 B 3.3.1-49 84 DRR 20-0400 08/18/20 B 3.3.1-50 84 DRR 20-0400 08/18/20 B 3.3.1-51 84 DRR 20-0400 08/18/20 B 3.3.1-52 84 DRR 20-0400 08/18/20 B 3.3.1-53 84 DRR 20-0400 08/18/20 B 3.3.1-54 84 DRR 20-0400 08/18/20 B 3.3.1-55 84 DRR 20-0400 08/18/20 B 3.3.1-56 84 DRR 20-0400 08/18/20 B 3.3.1-57 84 DRR 20-0400 08/18/20 B 3.3.1-58 84 DRR 20-0400 08/18/20 B 3.3.1-59 84 DRR 20-0400 08/18/20 B 3.3.1-60 84 DRR 20-0400 08/18/20 B 3.3.1-61 84 DRR 20-0400 08/18/20 B 3.3.1-62 84 DRR 20-0400 08/18/20 B 3.3.2-1 84 DRR 20-0400 08/18/20 B 3.3.2-2 0 Amend. No. 123 12/18/99 B 3.3.2-3 0 Amend. No. 123 12/18/99 B 3.3.2-4 0 Amend. No. 123 12/18/99 B 3.3.2-5 0 Amend. No. 123 12/18/99 B 3.3.2-6 7 DRR 01-0474 5/1/01 B 3.3.2-7 0 Amend. No. 123 12/18/99 B 3.3.2-8 0 Amend. No. 123 12/18/99 B 3.3.2-9 0 Amend. No. 123 12/18/99 B 3.3.2-10 0 Amend. No. 123 12/18/99 B 3.3.2-11 0 Amend. No. 123 12/18/99 B 3.3.2-12 81 DRR 19-1027 10/28/19 B 3.3.2-13 0 Amend. No. 123 12/18/99 B 3.3.2-14 2 DRR 00-0147 4/24/00 B 3.3.2-15 0 Amend. No. 123 12/18/99 B 3.3.2-16 0 Amend. No. 123 12/18/99 B 3.3.2-17 0 Amend. No. 123 12/18/99 B 3.3.2-18 0 Amend. No. 123 12/18/99 B 3.3.2-19 37 DRR 08-0503 4/8/08 B 3.3.2-20 37 DRR 08-0503 4/8/08 B 3.3.2-21 37 DRR 08-0503 4/8/08 B 3.3.2-22 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 iv Revision 88

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.2-23 37 DRR 08-0503 4/8/08 B 3.3.2-24 39 DRR 08-1096 8/28/08 B 3.3.2-25 39 DRR 08-1096 8/28/08 B 3.3.2-26 39 DRR 08-1096 8/28/08 B 3.3.2-27 37 DRR 08-0503 4/8/08 B 3.3.2-28 84 DRR 20-0400 08/18/20 B 3.3.2-29 0 Amend. No. 123 12/18/99 B 3.3.2-30 0 Amend. No. 123 12/18/99 B 3.3.2-31 52 DRR 11-0724 4/11/11 B 3.3.2-32 52 DRR 11-0724 4/11/11 B 3.3.2-33 0 Amend. No. 123 12/18/99 B 3.3.2-34 0 Amend. No. 123 12/18/99 B 3.3.2-35 20 DRR 04-1533 2/16/05 B 3.3.2-36 20 DRR 04-1533 2/16/05 B 3.3.2-37 20 DRR 04-1533 2/16/05 B 3.3.2-38 20 DRR 04-1533 2/16/05 B 3.3.2-39 25 DRR 06-0800 5/18/06 B 3.3.2-40 20 DRR 04-1533 2/16/05 B 3.3.2-41 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-42 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-43 20 DRR 04-1533 2/16/05 B 3.3.2-44 20 DRR 04-1533 2/16/05 B 3.3.2-45 20 DRR 04-1533 2/16/05 B 3.3.2-46 54 DRR 11-2394 11/16/11 B 3.3.2-47 43 DRR 09-1416 9/2/09 B 3.3.2-48 37 DRR 08-0503 4/8/08 B 3.3.2-49 20 DRR 04-1533 2/16/05 B 3.3.2-50 20 DRR 04-1533 2/16/05 B 3.3.2-51 43 DRR 09-1416 9/2/09 B 3.3.2-52 43 DRR 09-1416 9/2/09 B 3.3.2-53 43 DRR 09-1416 9/2/09 B 3.3.2-54 43 DRR 09-1416 9/2/09 B 3.3.2-55 43 DRR 09-1416 9/2/09 B 3.3.2-56 43 DRR 09-1416 9/2/09 B 3.3.2-57 43 DRR 09-1416 9/2/09 B 3.3.3-1 0 Amend. No. 123 12/18/99 B 3.3.3-2 5 DRR 00-1427 10/12/00 B 3.3.3-3 0 Amend. No. 123 12/18/99 B 3.3.3-4 0 Amend. No. 123 12/18/99 B 3.3.3-5 0 Amend. No. 123 12/18/99 B 3.3.3-6 8 DRR 01-1235 9/19/01 B 3.3.3-7 21 DRR 05-0707 4/20/05 B 3.3.3-8 81 DRR 19-1027 10/28/19 B 3.3.3-9 8 DRR 01-1235 9/19/01 B 3.3.3-10 19 DRR 04-1414 10/12/04 B 3.3.3-11 19 DRR 04-1414 10/12/04 B 3.3.3-12 21 DRR 05-0707 4/20/05 B 3.3.3-13 21 DRR 05-0707 4/20/05 B 3.3.3-14 8 DRR 01-1235 9/19/01 B 3.3.3-15 8 DRR 01-1235 9/19/01 B 3.3.4-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 v Revision 88

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.4-2 9 DRR 02-1023 2/28/02 B 3.3.4-3 15 DRR 03-0860 7/10/03 B 3.3.4-4 19 DRR 04-1414 10/12/04 B 3.3.4-5 1 DRR 99-1624 12/18/99 B 3.3.4-6 9 DRR 02-0123 2/28/02 B 3.3.5-1 88 DRR 21-0591 4/28/21 B 3.3.5-2 88 DRR 21-0591 4/28/21 B 3.3.5-3 1 DRR 99-1624 12/18/99 B 3.3.5-4 1 DRR 99-1624 12/18/99 B 3.3.5-5 0 Amend. No. 123 12/18/99 B 3.3.5-6 88 DRR 21-0591 4/28/21 B 3.3.5-7 88 DRR 21-0591 4/28/21 B 3.3.6-1 81 DRR 19-1027 10/28/19 B 3.3.6-2 81 DRR 19-1027 10/28/19 B 3.3.6-3 0 Amend. No. 123 12/18/99 B 3.3.6-4 0 Amend. No. 123 12/18/99 B 3.3.6-5 0 Amend. No. 123 12/18/99 B 3.3.6-6 0 Amend. No. 123 12/18/99 B 3.3.6-7 81 DRR 19-1027 10/28/19 B 3.3.7-1 81 DRR 19-1027 10/28/19 B 3.3.7-2 81 DRR 19-1027 10/28/19 B 3.3.7-3 57 DRR 13-0006 1/16/13 B 3.3.7-4 0 Amend. No. 123 12/18/99 B 3.3.7-5 0 Amend. No. 123 12/18/99 B 3.3.7-6 57 DRR 13-0006 1/16/13 B 3.3.7-7 0 Amend. No. 123 12/18/99 B 3.3.7-8 81 DRR 19-1027 10/28/19 B 3.3.8-1 84 DRR 20-0400 8/18/20 B 3.3.8-2 0 Amend. No. 123 12/18/99 B 3.3.8-3 57 DRR 13-0006 1/16/13 B 3.3.8-4 57 DRR 13-0006 1/16/13 B 3.3.8-5 0 Amend. No. 123 12/18/99 B 3.3.8-6 24 DRR 06-0051 2/28/06 B 3.3.8-7 81 DRR 19-1027 10/28/19 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1-1 84 DRR 20-0400 08/18/20 B 3.4.1-2 84 DRR 20-0400 08/18/20 B 3.4.1-3 10 DRR 02-0411 4/5/02 B 3.4.1-4 0 Amend. No. 123 12/18/99 B 3.4.1-5 0 Amend. No. 123 12/18/99 B 3.4.1-6 84 DRR 20-0400 08/18/20 B 3.4.2-1 0 Amend. No. 123 12/18/99 B 3.4.2-2 0 Amend. No. 123 12/18/99 B 3.4.2-3 0 Amend. No. 123 12/18/99 B 3.4.3-1 67 DRR 15-0116 2/10/15 B 3.4.3-2 0 Amend. No. 123 12/18/99 B 3.4.3-3 0 Amend. No. 123 12/18/99 B 3.4.3-4 0 Amend. No. 123 12/18/99 B 3.4.3-5 0 Amend. No. 123 12/18/99 B 3.4.3-6 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 vi Revision 88

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.3-7 0 Amend. No. 123 12/18/99 B 3.4.4-1 0 Amend. No. 123 12/18/99 B 3.4.4-2 29 DRR 06-1984 10/17/06 B 3.4.4-3 0 Amend. No. 123 12/18/99 B 3.4.5-1 0 Amend. No. 123 12/18/99 B 3.4.5-2 53 DRR 11-1513 7/18/11 B 3.4.5-3 29 DRR 06-1984 10/17/06 B 3.4.5-4 0 Amend. No. 123 12/18/99 B 3.4.5-5 12 DRR 02-1062 9/26/02 B 3.4.5-6 12 DRR 02-1062 9/26/02 B 3.4.6-1 53 DRR 11-1513 7/18/11 B 3.4.6-2 72 DRR 15-1918 10/26/15 B 3.4.6-3 12 DRR 02-1062 9/26/02 B 3.4.6-4 72 DRR 15-1918 10/26/15 B 3.4.6-5 75 DRR 16-1909 10/26/16 B 3.4.6-6 75 DRR 16-1909 10/26/16 B 3.4.7-1 12 DRR 02-1062 9/26/02 B 3.4.7-2 17 DRR 04-0453 5/26/04 B 3.4.7-3 72 DRR 15-1918 10/26/15 B 3.4.7-4 42 DRR 09-1009 7/16/09 B 3.4.7-5 72 DRR 15-1918 10/26/15 B 3.4.7-6 75 DRR 16-1909 10/26/16 B 3.4.8-1 53 DRR 11-1513 7/18/11 B 3.4.8-2 72 DRR 15-1918 10/26/15 B 3.4.8-3 42 DRR 09-1009 7/16/09 B 3.4.8-4 75 DRR 16-1909 10/26/16 B 3.4.8-5 72 DRR 15-1918 10/26/15 B 3.4.9-1 0 Amend. No. 123 12/18/99 B 3.4.9-2 0 Amend. No. 123 12/18/99 B 3.4.9-3 0 Amend. No. 123 12/18/99 B 3.4.9-4 0 Amend. No. 123 12/18/99 B 3.4.10-1 5 DRR 00-1427 10/12/00 B 3.4.10-2 5 DRR 00-1427 10/12/00 B 3.4.10-3 0 Amend. No. 123 12/18/99 B 3.4.10-4 32 DRR 07-0139 2/7/07 B 3.4.11-1 0 Amend. No. 123 12/18/99 B 3.4.11-2 1 DRR 99-1624 12/18/99 B 3.4.11-3 19 DRR 04-1414 10/12/04 B 3.4.11-4 0 Amend. No. 123 12/18/99 B 3.4.11-5 1 DRR 99-1624 12/18/99 B 3.4.11-6 0 Amend. No. 123 12/18/99 B 3.4.11-7 32 DRR 07-0139 2/7/07 B 3.4.12-1 61 DRR 14-0346 2/27/14 B 3.4.12-2 61 DRR 14-0346 2/27/14 B 3.4.12-3 0 Amend. No. 123 12/18/99 B 3.4.12-4 61 DRR 14-0346 2/27/14 B 3.4.12-5 61 DRR 14-0346 2/27/14 B 3.4.12-6 56 DRR 12-1792 11/7/12 B 3.4.12-7 61 DRR 14-0346 2/27/14 B 3.4.12-8 1 DRR 99-1624 12/18/99 B 3.4.12-9 56 DRR 12-1792 11/7/12 Wolf Creek - Unit 1 vii Revision 88

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.12-10 0 Amend. No. 123 12/18/99 B 3.4.12-11 61 DRR 14-0346 2/27/14 B 3.4.12-12 32 DRR 07-0139 2/7/07 B 3.4.12-13 0 Amend. No. 123 12/18/99 B 3.4.12-14 32 DRR 07-0139 2/7/07 B 3.4.13-1 0 Amend. No. 123 12/18/99 B 3.4.13-2 81 DRR 19-1027 10/28/19 B 3.4.13-3 29 DRR 06-1984 10/17/06 B 3.4.13-4 35 DRR 07-1553 9/28/07 B 3.4.13-5 35 DRR 07-1553 9/28/07 B 3.4.13-6 81 DRR 19-1027 10/28/19 B 3.4.14-1 0 Amend. No. 123 12/18/99 B 3.4.14-2 0 Amend. No. 123 12/18/99 B 3.4.14-3 0 Amend. No. 123 12/18/99 B 3.4.14-4 0 Amend. No. 123 12/18/99 B 3.4.14-5 32 DRR 07-0139 2/7/07 B 3.4.14-6 32 DRR 07-0139 2/7/07 B 3.4.15-1 31 DRR 06-2494 12/13/06 B 3.4.15-2 31 DRR 06-2494 12/13/06 B 3.4.15-3 33 DRR 07-0656 5/1/07 B 3.4.15-4 33 DRR 07-0656 5/1/07 B 3.4.15-5 65 DRR 14-2146 9/30/14 B 3.4.15-6 31 DRR 06-2494 12/13/06 B 3.4.15-7 31 DRR 06-2494 12/13/06 B 3.4.15-8 31 DRR 06-2494 12/13/06 B 3.4.16-1 81 DRR 19-1027 10/28/19 B 3.4.16-2 84 DRR 20-0400 08/18/20 B 3.4.16-3 31 DRR 06-2494 12/13/06 B 3.4.16-4 31 DRR 06-2494 12/13/06 B 3.4.16-5 81 DRR 19-1027 10/28/19 B 3.4.17-1 29 DRR 06-1984 10/17/06 B 3.4.17-2 81 DRR 19-1027 10/28/19 B 3.4.17-3 52 DRR 11-0724 4/11/11 B 3.4.17-4 81 DRR 19-1027 10/28/19 B 3.4.17-5 57 DRR 13-0006 1/16/13 B 3.4.17-6 57 DRR 13-0006 1/16/13 B 3.4.17-7 81 DRR 19-1027 10/28/19 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1-1 0 Amend. No. 123 12/18/99 B 3.5.1-2 0 Amend. No. 123 12/18/99 B 3.5.1-3 73 DRR 15-2135 11/17/15 B 3.5.1-4 73 DRR 15-2135 11/17/15 B 3.5.1-5 1 DRR 99-1624 12/18/99 B 3.5.1-6 1 DRR 99-1624 12/18/99 B 3.5.1-7 71 DRR 15-1528 7/30/15 B 3.5.1-8 1 DRR 99-1624 12/18/99 B 3.5.2-1 84 DRR 20-0400 08/18/20 B 3.5.2-2 0 Amend. No. 123 12/18/99 B 3.5.2-3 0 Amend. No. 123 12/18/99 B 3.5.2-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 viii Revision 88

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TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

B 3.5.2-5 72 DRR 15-1918 10/26/15 B 3.5.2-6 42 DRR 09-1009 7/16/09 B 3.5.2-7 42 DRR 09-1009 7/16/09 B 3.5.2-8 72 DRR 15-1918 10/26/15 B 3.5.2-9 75 DRR 16-1909 10/26/16 B 3.5.2-10 80 DRR 19-0524 5/30/19 B 3.5.2-11 72 DRR 15-1918 10/26/15 B 3.5.2-12 72 DRR 15-1918 10/26/15 B 3.5.3-1 56 DRR 12-1792 11/7/12 B 3.5.3-2 72 DRR 15-1918 10/26/15 B 3.5.3-3 56 DRR 12-1792 11/7/12 B 3.5.3-4 56 DRR 12-1792 11/7/12 B 3.5.4-1 0 Amend. No. 123 12/18/99 B 3.5.4-2 0 Amend. No. 123 12/18/99 B 3.5.4-3 0 Amend. No. 123 12/18/99 B 3.5.4-4 0 Amend. No. 123 12/18/99 B 3.5.4-5 0 Amend. No. 123 12/18/99 B 3.5.4-6 26 DRR 06-1350 7/24/06 B 3.5.5-1 21 DRR 05-0707 4/20/05 B 3.5.5-2 21 DRR 05-0707 4/20/05 B 3.5.5-3 2 Amend. No. 132 4/24/00 B 3.5.5-4 21 DRR 05-0707 4/20/05 TAB - B 3.6 CONTAINMENT SYSTEMS B 3.6.1-1 0 Amend. No. 123 12/18/99 B 3.6.1-2 81 DRR 19-1027 10/28/19 B 3.6.1-3 0 Amend. No. 123 12/18/99 B 3.6.1-4 87 DRR 21-0359 3/25/21 B 3.6.2-1 81 DRR 19-1027 10/28/19 B 3.6.2-2 0 Amend. No. 123 12/18/99 B 3.6.2-3 0 Amend. No. 123 12/18/99 B 3.6.2-4 0 Amend. No. 123 12/18/99 B 3.6.2-5 0 Amend. No. 123 12/18/99 B 3.6.2-6 0 Amend. No. 123 12/18/99 B 3.6.2-7 0 Amend. No. 123 12/18/99 B 3.6.3-1 0 Amend. No. 123 12/18/99 B 3.6.3-2 84 DRR 20-0400 08/18/20 B 3.6.3-3 81 DRR 19-1027 10/28/19 B 3.6.3-4 49 DRR 11-0014 1/31/11 B 3.6.3-5 49 DRR 11-0014 1/31/11 B 3.6.3-6 49 DRR 11-0014 1/31/11 B 3.6.3-7 41 DRR 09-0288 3/20/09 B 3.6.3-8 36 DRR 08-0255 3/11/08 B 3.6.3-9 36 DRR 08-0255 3/11/08 B 3.6.3-10 8 DRR 01-1235 9/19/01 B 3.6.3-11 36 DRR 08-0255 3/11/08 B 3.6.3-12 36 DRR 08-0255 3/11/08 B 3.6.3-13 50 DRR 11-0449 3/9/11 B 3.6.3-14 36 DRR 08-0255 3/11/08 B 3.6.3-15 39 DRR 08-1096 8/28/08 B 3.6.3-16 39 DRR 08-1096 8/28/08 Wolf Creek - Unit 1 ix Revision 88

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TAB - B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.3-17 36 DRR 08-0255 3/11/08 B 3.6.3-18 36 DRR 08-0255 3/11/08 B 3.6.3-19 36 DRR 08-0255 3/11/08 B 3.6.4-1 39 DRR 08-1096 8/28/08 B 3.6.4-2 0 Amend. No. 123 12/18/99 B 3.6.4-3 0 Amend. No. 123 12/18/99 B 3.6.5-1 0 Amend. No. 123 12/18/99 B 3.6.5-2 37 DRR 08-0503 4/8/08 B 3.6.5-3 13 DRR 02-1458 12/03/02 B 3.6.5-4 0 Amend. No. 123 12/18/99 B 3.6.6-1 81 DRR 19-1027 10/28/19 B 3.6.6-2 63 DRR 14-1572 7/1/14 B 3.6.6-3 37 DRR 08-0503 4/8/08 B 3.6.6-4 81 DRR 19-1027 10/28/19 B 3.6.6-5 0 Amend. No. 123 12/18/99 B 3.6.6-6 18 DRR 04-1018 9/1/04 B 3.6.6-7 72 DRR 15-1918 10/26/15 B 3.6.6-8 80 DRR 19-0524 5/30/19 B 3.6.6-9 72 DRR 15-1918 10/26/15 B 3.6.6-10 75 DRR 16-1909 10/26/16 B 3.6.6.11 80 DRR 19-0524 5/30/19 B 3.6.7-1 0 Amend. No. 123 12/18/99 B 3.6.7-2 81 DRR 19-1027 10/28/19 B 3.6.7-3 81 DRR 19-1027 10/28/19 B 3.6.7-4 81 DRR 19-1027 10/28/19 B 3.6.7-5 42 DRR 09-1009 7/16/09 TAB - B 3.7 PLANT SYSTEMS B 3.7.1-1 0 Amend. No. 123 12/18/99 B 3.7.1-2 84 DRR 20-0400 08/18/20 B 3.7.1-3 0 Amend. No. 123 12/18/99 B 3.7.1-4 84 DRR 20-0400 08/18/20 B 3.7.1-5 84 DRR 20-0400 08/18/20 B 3.7.1-6 84 DRR 20-0400 08/18/20 B 3.7.2-1 44 DRR 09-1744 10/28/09 B 3.7.2-2 44 DRR 09-1744 10/28/09 B 3.7.2-3 44 DRR 09-1744 10/28/09 B 3.7.2-4 81 DRR 19-1027 10/28/19 B 3.7.2-5 44 DRR 09-1744 10/28/09 B 3.7.2-6 44 DRR 09-1744 10/28/09 B 3.7.2-7 44 DRR 09-1744 10/28/09 B 3.7.2-8 44 DRR 09-1744 10/28/09 B 3.7.2-9 44 DRR 09-1744 10/28/09 B 3.7.2-10 81 DRR 19-1027 10/28/19 B 3.7.2-11 44 DRR 09-1744 10/28/09 B 3.7.3-1 37 DRR 08-0503 4/8/08 B 3.7.3-2 50 DRR 11-0449 3/9/11 B 3.7.3-3 37 DRR 08-0503 4/8/08 B 3.7.3-4 37 DRR 08-0503 4/8/08 B 3.7.3-5 37 DRR 08-0503 4/8/08 B 3.7.3-6 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 x Revision 88

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.3-7 37 DRR 08-0503 4/8/08 B 3.7.3-8 37 DRR 08-0503 4/8/08 B 3.7.3-9 66 DRR 14-2329 11/6/14 B 3.7.3-10 66 DRR 14-2329 11/6/14 B 3.7.3-11 37 DRR 08-0503 4/8/08 B 3.7.4-1 1 DRR 99-1624 12/18/99 B 3.7.4-2 84 DRR 20-0400 08/18/20 B 3.7.4-3 19 DRR 04-1414 10/12/04 B 3.7.4-4 19 DRR 04-1414 10/12/04 B 3.7.4-5 84 DRR 20-0400 08/18/20 B 3.7.5-1 54 DRR 11-2394 11/16/11 B 3.7.5-2 54 DRR 11-2394 11/16/11 B 3.7.5-3 0 Amend. No. 123 12/18/99 B 3.7.5-4 85 DRR 20-0988 10/24/20 B 3.7.5-5 76 DRR 17-0343 2/21/17 B 3.7.5-6 85 DRR 20-0988 10/24/20 B 3.7.5-7 85 DRR 20-0988 10/24/20 B 3.7.5-8 85 DRR 20-0988 10/24/20 B 3.7.5-9 85 DRR 20-0988 10/24/20 B 3.7.5-10 85 DRR 20-0988 10/24/20 B 3.7.6-1 0 Amend. No. 123 12/18/99 B 3.7.6-2 0 Amend. No. 123 12/18/99 B 3.7.6-3 0 Amend. No. 123 12/18/99 B 3.7.7-1 0 Amend. No. 123 12/18/99 B 3.7.7-2 0 Amend. No. 123 12/18/99 B 3.7.7-3 0 Amend. No. 123 12/18/99 B 3.7.7-4 1 DRR 99-1624 12/18/99 B 3.7.8-1 0 Amend. No. 123 12/18/99 B 3.7.8-2 0 Amend. No. 123 12/18/99 B 3.7.8-3 0 Amend. No. 123 12/18/99 B 3.7.8-4 0 Amend. No. 123 12/18/99 B 3.7.8-5 0 Amend. No. 123 12/18/99 B 3.7.9-1 3 Amend. No. 134 7/14/00 B 3.7.9-2 3 Amend. No. 134 7/14/00 B 3.7.9-3 3 Amend. No. 134 7/14/00 B 3.7.9-4 3 Amend. No. 134 7/14/00 B 3.7.10-1 64 DRR 14-1822 8/28/14 B 3.7.10-2 81 DRR 19-1027 10/28/19 B 3.7.10-3 81 DRR 19-1027 10/28/19 B 3.7.10-4 81 DRR 19-1027 10/28/19 B 3.7.10-5 81 DRR 19-1027 10/28/19 B 3.7.10-6 57 DRR 13-0006 1/16/13 B 3.7.10-7 64 DRR 14-1822 8/28/14 B 3.7.10-8 81 DRR 19-1027 10/28/19 B 3.7.10-9 81 DRR 19-1027 10/28/19 B 3.7.11-1 0 Amend. No. 123 12/18/99 B 3.7.11-2 57 DRR 13-0006 1/16/13 B 3.7.11-3 63 DRR 14-1572 7/1/14 B 3.7.11-4 63 DRR 14-1572 7/1/14 B 3.7.12-1 0 Amend. No. 123 12/18/99 B 3.7.13-1 24 DRR 06-0051 2/28/06 Wolf Creek - Unit 1 xi Revision 88

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IMPLEMENTED (4)

TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.13-2 81 DRR 19-1027 10/28/19 B 3.7.13-3 81 DRR 19-1027 10/28/19 B 3.7.13-4 81 DRR 19-1027 10/28/19 B 3.7.13-5 81 DRR 19-1027 10/28/19 B 3.7.13-6 81 DRR 19-1027 10/28/19 B 3.7.13-7 81 DRR 19-1027 10/28/19 B 3.7.13-8 81 DRR 19-1027 10/28/19 B 3.7.14-1 0 Amend. No. 123 12/18/99 B 3.7.15-1 81 DRR 19-1027 10/28/19 B 3.7.15-2 81 DRR 19-1027 10/28/19 B 3.7.15-3 81 DRR 19-1027 10/28/19 B 3.7.16-1 5 DRR 00-1427 10/12/00 B 3.7.16-2 23 DRR 05-1995 9/28/05 B 3.7.16-3 5 DRR 00-1427 10/12/00 B 3.7.17-1 7 DRR 01-0474 5/1/01 B 3.7.17-2 7 DRR 01-0474 5/1/01 B 3.7.17-3 5 DRR 00-1427 10/12/00 B 3.7.18-1 81 DRR 19-1027 10/28/19 B 3.7.18-2 81 DRR 19-1027 10/28/19 B 3.7.18-3 81 DRR 19-1027 10/28/19 B 3.7.19-1 44 DRR 09-1744 10/28/09 B 3.7.19-2 54 DRR 11-2394 11/16/11 B 3.7.19-3 54 DRR 11-2394 11/16/11 B 3.7.19-4 61 DRR 14-0346 2/27/14 B 3.7.19-5 61 DRR 14-0346 2/27/14 B 3.7.19-6 54 DRR 11-2394 11/16/11 B 3.7.19-7 54 DRR 11-2394 11/16/11 B 3.7.20-1 79 DRR 18-1579 10/22/18 B 3.7.20-2 79 DRR 18-1579 10/22/18 B 3.7.20-3 85 DRR 20-0988 10/24/20 B 3.7.20-4 79 DRR 18-1579 10/22/18 B 3.7.20-5 79 DRR 18-1579 10/22/18 TAB - B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1-1 88 DRR 21-0591 4/28/21 B 3.8.1-2 88 DRR 21-0591 4/28/21 B 3.8.1-3 88 DRR 21-0951 4/28/21 B 3.8.1-4 86 DRR 21-0212 2/18/21 B 3.8.1-5 59 DRR 13-1524 6/26/13 B 3.8.1-6 25 DRR 06-0800 5/18/06 B 3.8.1-7 26 DRR 06-1350 7/24/06 B 3.8.1-8 35 DRR 07-1553 9/28/07 B 3.8.1-9 42 DRR 09-1009 7/16/09 B 3.8.1-10 39 DRR 08-1096 8/28/08 B 3.8.1-11 36 DRR 08-0255 3/11/08 B 3.8.1-12 75 DRR 16-1909 10/26/16 B 3.8.1-13 47 DRR 10-1089 6/16/10 B 3.8.1-14 47 DRR 10-1089 6/16/10 B 3.8.1-15 47 DRR 10-1089 6/16/10 B 3.8.1-16 26 DRR 06-1350 7/24/06 B 3.8.1-17 26 DRR 06-1350 7/24/06 Wolf Creek - Unit 1 xii Revision 88

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.1-18 59 DRR 13-1524 6/26/13 B 3.8.1-19 26 DRR 06-1350 7/24/06 B 3.8.1-20 26 DRR 06-1350 7/24/06 B 3.8.1-21 33 DRR 07-0656 5/1/07 B 3.8.1-22 33 DRR 07-0656 5/1/07 B 3.8.1-23 74 DRR 16-1182 7/7/16 B 3.8.1-24 74 DRR 16-1182 7/7/16 B 3.8.1-25 74 DRR 16-1182 7/7/16 B 3.8.1-26 74 DRR 16-1182 7/7/16 B 3.8.1-27 74 DRR 16-1182 7/7/16 B 3.8.1-28 74 DRR 16-1182 7/7/16 B 3.8.1-29 74 DRR 16-1182 7/7/16 B 3.8.1-30 74 DRR 16-1182 7/7/16 B 3.8.1-31 74 DRR 16-1182 7/7/16 B 3.8.1-32 74 DRR 16-1182 7/7/16 B 3.8.1-33 74 DRR 16-1182 7/7/16 B 3.8.1-34 74 DRR 16-1182 7/7/16 B 3.8.2-1 57 DRR 13-0006 1/16/13 B 3.8.2-2 0 Amend. No. 123 12/18/99 B 3.8.2-3 86 DRR 21-0212 2/18/21 B 3.8.2-4 88 DRR 21-0591 4/28/21 B 3.8.2-5 57 DRR 13-0006 1/16/13 B 3.8.2-6 57 DRR 13-0006 1/16/13 B 3.8.2-7 57 DRR 13-0006 1/16/13 B 3.8.3-1 1 DRR 99-1624 12/18/99 B 3.8.3-2 0 Amend. No. 123 12/18/99 B 3.8.3-3 0 Amend. No. 123 12/18/99 B 3.8.3-4 1 DRR 99-1624 12/18/99 B 3.8.3-5 0 Amend. No. 123 12/18/99 B 3.8.3-6 0 Amend. No. 123 12/18/99 B 3.8.3-7 12 DRR 02-1062 9/26/02 B 3.8.3-8 1 DRR 99-1624 12/18/99 B 3.8.3-9 0 Amend. No. 123 12/18/99 B 3.8.4-1 0 Amend. No. 123 12/18/99 B 3.8.4-2 0 Amend. No. 123 12/18/99 B 3.8.4-3 0 Amend. No. 123 12/18/99 B 3.8.4-4 0 Amend. No. 123 12/18/99 B 3.8.4-5 50 DRR 11-0449 3/9/11 B 3.8.4-6 50 DRR 11-0449 3/9/11 B 3.8.4-7 6 DRR 00-1541 3/13/01 B 3.8.4-8 0 Amend. No. 123 12/18/99 B 3.8.4-9 2 DRR 00-0147 4/24/00 B 3.8.5-1 57 DRR 13-0006 1/16/13 B 3.8.5-2 0 Amend. No. 123 12/18/99 B 3.8.5-3 57 DRR 13-0006 1/16/13 B 3.8.5-4 57 DRR 13-0006 1/16/13 B 3.8.5-5 57 DRR 13-0006 1/16/13 B 3.8.6-1 0 Amend. No. 123 12/18/99 B 3.8.6-2 0 Amend. No. 123 12/18/99 B 3.8.6-3 0 Amend. No. 123 12/18/99 B 3.8.6-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiii Revision 88

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.6-5 0 Amend. No. 123 12/18/99 B 3.8.6-6 0 Amend. No. 123 12/18/99 B 3.8.7-1 69 DRR 15-0493 3/26/15 B 3.8.7-2 69 DRR 15-0493 3/26/15 B 3.8.7-3 69 DRR 15-0493 3/26/15 B 3.8.7-4 0 Amend. No. 123 12/18/99 B 3.8.8-1 57 DRR 13-0006 1/16/13 B 3.8.8-2 0 Amend. No. 123 12/18/99 B 3.8.8-3 69 DRR 15-0493 3/26/15 B 3.8.8-4 57 DRR 13-0006 1/16/13 B 3.8.8-5 69 DRR 15-0493 3/26/15 B 3.8.9-1 54 DRR 11-2394 11/16/11 B 3.8.9-2 69 DRR 15-0493 3/26/15 B 3.8.9-3 54 DRR 11-2394 11/16/11 B 3.8.9-4 0 Amend. No. 123 12/18/99 B 3.8.9-5 69 DRR 15-0493 3/26/15 B 3.8.9-6 0 Amend. No. 123 12/18/99 B 3.8.9-7 0 Amend. No. 123 12/18/99 B 3.8.9-8 1 DRR 99-1624 12/18/99 B 3.8.9-9 0 Amend. No. 123 12/18/99 B 3.8.10-1 57 DRR 13-0006 1/16/13 B 3.8.10-2 0 Amend. No. 123 12/18/99 B 3.8.10-3 0 Amend. No. 123 12/18/99 B 3.8.10-4 57 DRR 13-0006 1/16/13 B 3.8.10-5 57 DRR 13-0006 1/16/13 B 3.8.10-6 57 DRR 13-0006 1/16/13 TAB - B 3.9 REFUELING OPERATIONS B 3.9.1-1 0 Amend. No. 123 12/18/99 B 3.9.1-2 19 DRR 04-1414 10/12/04 B 3.9.1-3 19 DRR 04-1414 10/12/04 B 3.9.1-4 19 DRR 04-1414 10/12/04 B 3.9.2-1 0 Amend. No. 123 12/18/99 B 3.9.2-2 0 Amend. No. 123 12/18/99 B 3.9.2-3 0 Amend. No. 123 12/18/99 B 3.9.3-1 68 DRR 15-0248 2/26/15 B 3.9.3-2 68 DRR 15-0248 2/26/15 B 3.9.3-3 51 DRR 11-0664 3/21/11 B 3.9.3-4 68 DRR 15-0248 2/26/15 B 3.9.4-1 81 DRR 19-1027 10/28/19 B 3.9.4-2 13 DRR 02-1458 12/03/02 B 3.9.4-3 81 DRR 19-1027 10/28/19 B 3.9.4-4 23 DRR 05-1995 9/28/05 B 3.9.4-5 33 DRR 07-0656 5/1/07 B 3.9.4-6 23 DRR 05-1995 9/28/05 B 3.9.5-1 0 Amend. No. 123 12/18/99 B 3.9.5-2 72 DRR 15-1918 10/26/15 B 3.9.5-3 32 DRR 07-0139 2/7/07 B 3.9.5-4 75 DRR 16-1909 10/26/16 B 3.9.5-5 75 DRR 16-1909 10/26/16 B 3.9.6-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiv Revision 88

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TAB - B 3.9 REFUELING OPERATIONS (continued)

B 3.9.6-2 72 DRR 15-1918 10/26/15 B 3.9.6-3 42 DRR 09-1009 7/16/09 B 3.9.6-4 72 DRR 15-1918 10/26/15 B 3.9.6-5 75 DRR 16-1909 10/26/16 B 3.9.7-1 81 DRR 19-1027 10/28/19 B 3.9.7-2 81 DRR 19-1027 10/28/19 B 3.9.7-3 81 DRR 19-1027 10/28/19 Note 1 The page number is listed on the center of the bottom of each page.

Note 2 The revision number is listed in the lower right hand corner of each page. The Revision number will be page specific.

Note 3 The change document will be the document requesting the change. Amendment No.

123 issued the improved Technical Specifications and associated Bases which affected each page. The NRC has indicated that Bases changes will not be issued with License Amendments. Therefore, the change document should be a DRR number in accordance with AP 26A-002.

Note 4 The date effective or implemented is the date the Bases pages are issued by Document Control.

Wolf Creek - Unit 1 xv Revision 88

ENCLOSURE Page 51 of 264 LOP DG Start Instrumentation B 3.3.5 B 3.3 INSTRUMENTATION B 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation BASES BACKGROUND The DGs provide a source of emergency power when offsite power is either unavailable or is insufficiently stable to allow safe unit operation.

There are two sets of undervoltage/degraded voltage protection circuits, one for each 4.16 kV NB system bus. Each set consists of a loss of voltage and degraded voltage Function. The undervoltage/degraded voltage protection circuits are described in USAR Section 8.3.1.1.3 (Ref. 1).

Upon recognition of a loss of voltage at the 4.16 kV ESF buses, a logic signal generated by load shedder and emergency load sequencer (LSELS) initiates the following:

a) Trip the 4.16 kV preferred normal and alternate bus feeder breakers to remove the deficient power source to protect Class 1E equipment from damage; b) Shed all loads from the bus except the Class 1E 480 Vac load centers and centrifugal charging pumps to prepare the buses for re-energization by the LSELS; and c) Generate an LOP DG start signal.

Upon detection of a degraded voltage condition, LSELS initiates a logic signal which serves only to trip the 4.16 kV ESF bus normal and alternate feeder breakers. The undervoltage relays detect an undervoltage condition and the same initiation signals as described above are actuated.

Four instantaneous undervoltage relays with an associated time delay are provided for each 4.16 kV Class 1E NB system bus for detecting a loss of bus voltage. The outputs are combined in a two-out-of-four logic to generate an LOP signal if the voltage is below approximately 78% for 1 second (nominal delay). The time delay prevents undesirable trips arising from transient undervoltage conditions.

Four potential transformers provide input to four degraded voltage bistables with associated time delays for each 4.16 kV Class 1E system bus for detecting a sustained degraded voltage condition. Once the bistable has actuated, a timer in the LSELS circuitry provides an 8 second time delay to avoid false actuation on large motor starts other than an Wolf Creek - Unit 1 B 3.3.5-1 Revision 88

ENCLOSURE Page 52 of 264 LOP DG Start Instrumentation B 3.3.5 BASES BACKGROUND RCP. There are four of these 8-second timers per bus, one for each (continued) degraded voltage channel. The bistable outputs are then combined in a two-out-of-four logic to generate a degraded voltage signal if the voltage is below approximately 92%. Once the two-out-of four logic is satisfied, contacts in the bus feeder breaker trip circuits closed to arm the tripping circuitry. The degraded voltage time delay with a safety injection signal (SIS) present is 8.0 seconds. If an SIS were to occur concurrently with or after the arming of the tripping circuitry, the bus feeder breaker would open immediately, a bus undervoltage would be sensed, and a LOP signal would be generated. Should the degraded voltage condition occur in a non-accident condition (no SIS present), an additional 48 second time delay is provided. These time delays are specific to the feeder breakers (2 per bus). If the degraded voltage is not alleviated in the overall 56 seconds (nominal delay), the bus feeder breaker is tripped.

OPERABILITY of LSELS is addressed in LCO 3.8.1, AC Sources -

Operating, And LCO 3.8.2, AC Sources - Shutdown.

Trip Setpoints and Allowable Values The Trip Setpoints used in the relays are based on References 1 and 2.

The selection of these Trip Setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account.

The actual nominal Trip Setpoint entered into the relays is normally still more conservative than that required by the Allowable Value. The Trip Setpoints are the nominal value at which the bistables are set. Any bistable is considered to be properly adjusted when the as left value is within the two-sided tolerance band for channel accuracy. If the measured setpoint does not exceed the Allowable Value, the relay is considered OPERABLE.

Setpoints adjusted in accordance with the Allowable Value ensure that the consequences of accidents will be acceptable, provided the unit is operated from within the LCOs at the onset of the accident and that the equipment functions as designed.

Allowable Values and/or nominal Trip Setpoints are specified for each Function in SR 3.3.5.3. Nominal Trip Setpoints are also specified in the unit specific setpoint calculations. The nominal setpoints are selected to ensure that the setpoint measured by the surveillance procedure does not exceed the Allowable Value if the relay is performing as required. If the Wolf Creek - Unit 1 B 3.3.5-2 Revision 88

ENCLOSURE Page 53 of 264

ENCLOSURE Page 54 of 264

ENCLOSURE Page 55 of 264 AC Sources - Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources - Operating BASES BACKGROUND The unit Class 1E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources, normal and alternate from the redundant Engineered Safety Feature (ESF) transformers), and the onsite standby power sources (Train A and Train B diesel generators (DGs)). As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the ESF systems.

The onsite Class 1E AC Distribution System is divided into redundant load groups (trains) so that the loss of any one group does not prevent the minimum safety functions from being performed. Each train has connections to its preferred offsite power source and a single DG.

Offsite power is supplied to the unit switchyard from the transmission network by three transmission lines. From the switchyard, two electrically and physically separated circuits provide AC power, through the ESF transformers, to the 4.16 kV ESF buses. A detailed description of the offsite power network and the circuits to the Class 1E ESF buses is found in the USAR, Chapter 8 (Ref. 2).

An offsite circuit consists of all breakers, transformers, voltage regulating tap changers (when installed), switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class 1E ESF bus(es).

Certain required unit loads are returned to service in a predetermined sequence in order to prevent overloading the transformer supplying offsite power to the onsite Class 1E Distribution System. Within 1 minute after the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safe condition are returned to service via the load sequencer.

The onsite standby power source for each 4.16 kV ESF bus is a dedicated DG. DGs A and B are dedicated to ESF buses NB01 and NB02, respectively. The DG starts automatically on a safety injection (SI) signal or on an ESF bus undervoltage signal. A degraded voltage signal produces an undervoltage condition by opening the normal and alternate feeder breakers to the bus(es) experiencing degraded voltage. Both signals are initiated from the load shedder and emergency load sequencer (LSELS). OPERABILITY of the undervoltage and degraded voltage instrumentation functions are addressed in LCO 3.3.5, Loss of Wolf Creek - Unit 1 B 3.8.1-1 Revision 88

ENCLOSURE Page 56 of 264 AC Sources - Operating B 3.8.1 BASES BACKGROUND Power (LOP) Diesel Generator (DG) Start Instrumentation. After the DG (continued) has started, it will automatically tie to its respective bus after offsite power is tripped as a consequence of ESF bus undervoltage or degraded voltage, independent of or coincident with an SI signal. The DGs will also start and operate in the standby mode without tying to the ESF bus on an SI signal alone. Following the trip of offsite power, a LSELS strips non-essential loads from the ESF bus. When the DG is tied to the ESF bus, essential loads are then sequentially connected to its respective ESF bus by the load sequencer. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading the DG by automatic load application.

In the event of a loss of preferred power, the ESF electrical loads are automatically connected to the DGs in sufficient time to provide for safe reactor shutdown and to mitigate the consequences of a Design Basis Accident (DBA) such as a loss of coolant accident (LOCA).

Certain required unit loads are returned to service in a predetermined sequence in order to prevent overloading the DG in the process. Within 1 minute after the initiating signal is received, all loads needed to recover the unit or maintain it in a safe condition are returned to service.

Ratings for Train A and Train B DGs satisfy the requirements of Regulatory Guide 1.9 (Ref. 3). The continuous service rating of each DG is 6201 kW with 10% overload permissible for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. The ESF loads that are powered from the 4.16 kV ESF buses are listed in Reference 2.

APPLICABLE The initial conditions of DBA and transient analyses in the USAR, SAFETY ANALYSES Chapter 6 (Ref. 4) and Chapter 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded.

These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS);

and Section 3.6, Containment Systems.

The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the Accident analyses and is based upon meeting the design basis of the unit. This results in maintaining at least one train of the onsite or offsite AC sources OPERABLE during Accident conditions in the event of:

a. An assumed loss of all offsite power or all onsite AC power; and Wolf Creek - Unit 1 B 3.8.1-2 Revision 88

ENCLOSURE Page 57 of 264 AC Sources - Operating B 3.8.1 BASES APPLICABLE b. A worst case single failure.

SAFETY ANALYSES (continued) The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two qualified circuits between the offsite transmission network and the onsite Class 1E Electrical Power System, separate and independent DGs for each train, and redundant LSELS for each train ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (AOO) or a postulated DBA.

Each offsite circuit must be capable of maintaining rated frequency and voltage, and accepting required loads during an accident, while connected to the ESF buses.

One offsite circuit consists of the #7 transformer feeding through the 13-48 breaker power the ESF transformer XNB01, which, in turn powers the NB01 bus through its normal feeder breaker. Transformer XNB01 may also be powered from the SL-7 supply through the 13-8 breaker provided the offsite 69 KV line is not connected to the 345 kV system.

The lineup for transformers 4, 5, and 6 is an alternative to the #7 transformer. However, the alternate lineup does not have open phase detection instrumentation. The lineup with the 4, 5, and 6 transformers should only be used as an alternative in a risk analyzed condition and is not intended to be used for convenience of maintenance. The #7 transformer lineup should be restored as quickly as possible. The offsite circuit energizing NB01 is considered inoperable when NB01 is only energized from the transmission network through the 345-50 and 345-60 main generator breakers. For this configuration, switchyard breakers 345-120 and 345-80 are open.

Another offsite circuit consists of the startup transformer feeding through breaker PA201 powering the ESF transformer XNB02, which, in turn powers the NB02 bus through its normal feeder breaker.

ESF transformer XNB02 has an automatic load tap changer (LTC). In order for the ESF transformer to be OPERABLE, the LTC must be capable of meeting specified safety design functions when in automatic mode (ref. calculation XX-E-006), or at tap setting 3L or higher when in the fixed tap (manual) mode.

Each DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage. This will be accomplished within 12 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be Wolf Creek - Unit 1 B 3.8.1-3 Revision 88

ENCLOSURE Page 58 of 264 AC Sources - Shutdown B 3.8.2 BASES LCO of the startup transformer feeding through breaker PA201 powering the (continued) ESF transformer XNB02, which in turn powers the NB02 bus through its normal feeder breaker. The ESF transformer XNB02 has an automatic load tap changer (LTC). The LTC is discussed in the Bases for TS 3.8.1.

The DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective ESF bus on detection of bus undervoltage. This sequence must be accomplished within 12 seconds.

The DG must be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the ESF buses. These capabilities are required to be met from a variety of initial conditions such as DG in standby with the engine hot and DG in standby at ambient conditions.

Initiating an EDG start upon a detected under voltage or degraded voltage condition, tripping of nonessential loads, and proper sequencing of loads, is a required function of load shedder and emergency load sequencer (LSELS) and required for DG OPERABILITY. Only the shutdown sequencer on the train supported by the OPERABLE DG is required to be OPERABLE in MODES 5 and 6. In addition, the LSELS Automatic Test Indicator (ATI) is an installed testing aid and is not required to be OPERABLE to support the sequencer function. Absence of a functioning ATI does not render LSELS inoperable.

It is acceptable for one offsite circuit to be connected to more than one ESF bus through the normal or alternate feeder breakers with the loading limitations of calculation XX-E-006, "AC System Analysis," not being exceeded. Each offsite circuit can be manually aligned to supply power to the opposite NB bus.

APPLICABILITY The AC sources required to be OPERABLE in MODES 5 and 6, and during movement of irradiated fuel assemblies provide assurance that:

a. Systems to provide adequate coolant inventory makeup are available for the irradiated fuel assemblies in the core;

b. Systems needed to mitigate a fuel handling accident are available;

c. Systems necessary to mitigate the effects of events that can lead to core damage during shutdown are available; and

d. Instrumentation and control capability is available for monitoring and maintaining the unit in a cold shutdown condition or refueling condition.

The AC power requirements for MODES 1, 2, 3, and 4 are covered in LCO 3.8.1.

Wolf Creek - Unit 1 B 3.8.2-4 Revision 88

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TAB - Title Page Technical Specification Cover Page Title Page TAB - Table of Contents i 34 DRR 07-1057 7/10/07 ii 29 DRR 06-1984 10/17/06 iii 44 DRR 09-1744 10/28/09 TAB - B 2.0 SAFETY LIMITS (SLs)

B 2.1.1-1 0 Amend. No. 123 12/18/99 B 2.1.1-2 14 DRR 03-0102 2/12/03 B 2.1.1-3 14 DRR 03-0102 2/12/03 B 2.1.1-4 0 Amend. No. 123 2/12/03 B 2.1.2-1 84 DRR 20-0400 08/18/20 B 2.1.2-2 84 DRR 20-0400 08/18/20 B 2.1.2-3 81 DRR 19-1027 10/28/19 TAB - B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 81 DRR 19-1027 10/28/19 B 3.0-2 0 Amend. No. 123 12/18/99 B 3.0-3 81 DRR 19-1027 10/28/19 B 3.0-4 81 DRR 19-1027 10/28/19 B 3.0-5 81 DRR 19-1027 10/28/19 B 3.0-6 81 DRR 19-1027 10/28/19 B 3.0-7 81 DRR 19-1027 10/28/19 B 3.0-8 81 DRR 19-1027 10/28/19 B 3.0-9 81 DRR 19-1027 10/28/19 B 3.0-10 81 DRR 19-1027 10/28/19 B 3.0-11 81 DRR 19-1027 10/28/19 B 3.0-12 81 DRR 19-1027 10/28/19 B 3.0-13 81 DRR 19-1027 10/28/19 B 3.0-14 81 DRR 19-1027 10/28/19 B 3.0-15 81 DRR 19-1027 10/28/19 B 3.0-16 81 DRR 19-1027 10/28/19 B 3.0-17 81 DRR 19-1027 10/28/19 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1-1 0 Amend. No. 123 12/18/99 B 3.1.1-2 0 Amend. No. 123 12/18/99 B 3.1.1-3 0 Amend. No. 123 12/18/99 B 3.1.1-4 81 DRR 19-1027 10/28/19 B 3.1.1-5 89 DRR 21-0966 7/7/21 B 3.1.2-1 0 Amend. No. 123 12/18/99 B 3.1.2-2 0 Amend. No. 123 12/18/99 B 3.1.2-3 0 Amend. No. 123 12/18/99 B 3.1.2-4 0 Amend. No. 123 12/18/99 B 3.1.2-5 89 DRR 21-0966 7/7/21 B 3.1.3-1 0 Amend. No. 123 12/18/99 B 3.1.3-2 0 Amend. No. 123 12/18/99 B 3.1.3-3 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 i Revision 89

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TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)

B 3.1.3-4 0 Amend. No. 123 12/18/99 B 3.1.3-5 0 Amend. No. 123 12/18/99 B 3.1.3-6 0 Amend. No. 123 12/18/99 B 3.1.4-1 0 Amend. No. 123 12/18/99 B 3.1.4-2 0 Amend. No. 123 12/18/99 B 3.1.4-3 48 DRR 10-3740 12/28/10 B 3.1.4-4 0 Amend. No. 123 12/18/99 B 3.1.4-5 0 Amend. No. 123 12/18/99 B 3.1.4-6 48 DRR 10-3740 12/28/10 B 3.1.4-7 0 Amend. No. 123 12/18/99 B 3.1.4-8 89 DRR 21-0966 7/7/21 B 3.1.4-9 89 DRR 21-0966 7/7/21 B 3.1.5-1 0 Amend. No. 123 12/18/99 B 3.1.5-2 0 Amend. No. 123 12/18/99 B 3.1.5-3 0 Amend. No. 123 12/18/99 B 3.1.5-4 89 DRR 21-0966 7/7/21 B 3.1.6-1 0 Amend. No. 123 12/18/99 B 3.1.6-2 0 Amend. No. 123 12/18/99 B 3.1.6-3 0 Amend. No. 123 12/18/99 B 3.1.6-4 0 Amend. No. 123 12/18/99 B 3.1.6-5 89 DRR 21-0966 7/7/21 B 3.1.6-6 0 Amend. No. 123 12/18/99 B 3.1.7-1 0 Amend. No. 123 12/18/99 B 3.1.7-2 0 Amend. No. 123 12/18/99 B 3.1.7-3 48 DRR 10-3740 12/28/10 B 3.1.7-4 48 DRR 10-3740 12/28/10 B 3.1.7-5 48 DRR 10-3740 12/28/10 B 3.1.7-6 0 Amend. No. 123 12/18/99 B 3.1.8-1 0 Amend. No. 123 12/18/99 B 3.1.8-2 0 Amend. No. 123 12/18/99 B 3.1.8-3 15 DRR 03-0860 7/10/03 B 3.1.8-4 15 DRR 03-0860 7/10/03 B 3.1.8-5 89 DRR 21-0966 7/7/21 B 3.1.8-6 89 DRR 21-0966 7/7/21 B 3.1.9-1 84 DRR 20-0400 08/18/20 B 3.1.9-2 84 DRR 20-0400 08/18/20 B 3.1.9-3 84 DRR 20-0400 08/18/20 B 3.1.9-4 84 DRR 20-0400 08/18/20 B 3.1.9-5 89 DRR 21-0966 7/7/21 TAB - B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1-1 48 DRR 10-3740 12/28/10 B 3.2.1-2 0 Amend. No. 123 12/18/99 B 3.2.1-3 48 DRR 10-3740 12/28/10 B 3.2.1-4 48 DRR 10-3740 12/28/10 B 3.2.1-5 48 DRR 10-3740 12/28/10 B 3.2.1-6 48 DRR 10-3740 12/28/10 B 3.2.1-7 48 DRR 10-3740 12/28/10 B 3.2.1-8 89 DRR 21-0966 7/7/21 B 3.2.1-9 89 DRR 21-0966 7/7/21 B 3.2.1-10 70 DRR 15-0944 4/28/15 Wolf Creek - Unit 1 ii Revision 89

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TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued)

B 3.2.2-1 48 DRR 10-3740 12/28/10 B 3.2.2-2 0 Amend. No. 123 12/18/99 B 3.2.2-3 48 DRR 10-3740 12/28/10 B 3.2.2-4 48 DRR 10-3740 12/28/10 B 3.2.2-5 48 DRR 10-3740 12/28/10 B 3.2.2-6 89 DRR 21-0966 7/7/21 B 3.2.3-1 0 Amend. No. 123 12/18/99 B 3.2.3-2 0 Amend. No. 123 12/18/99 B 3.2.3-3 89 DRR 21-0966 7/7/21 B 3.2.4-1 0 Amend. No. 123 12/18/99 B 3.2.4-2 0 Amend. No. 123 12/18/99 B 3.2.4-3 48 DRR 10-3740 12/28/10 B 3.2.4-4 0 Amend. No. 123 12/18/99 B 3.2.4-5 48 DRR 10-3740 12/28/10 B 3.2.4-6 89 DRR 21-0966 7/7/21 B 3.2.4-7 89 DRR 21-0966 7/7/21 TAB - B 3.3 INSTRUMENTATION B 3.3.1-1 84 DRR 20-0400 08/18/20 B 3.3.1-2 0 Amend. No. 123 12/18/99 B 3.3.1-3 0 Amend. No. 123 12/18/99 B 3.3.1-4 0 Amend. No. 123 12/18/99 B 3.3.1-5 0 Amend. No. 123 12/18/99 B 3.3.1-6 0 Amend. No. 123 12/18/99 B 3.3.1-7 5 DRR 00-1427 10/12/00 B 3.3.1-8 0 Amend. No. 123 12/18/99 B 3.3.1-9 84 DRR 20-0400 08/18/20 B 3.3.1-10 84 DRR 20-0400 08/18/20 B 3.3.1-11 84 DRR 20-0400 08/18/20 B 3.3.1-12 84 DRR 20-0400 08/18/20 B 3.3.1-13 84 DRR 20-0400 08/18/20 B 3.3.1-14 84 DRR 20-0400 08/18/20 B 3.3.1-15 84 DRR 20-0400 08/18/20 B 3.3.1-16 84 DRR 20-0400 08/18/20 B 3.3.1-17 84 DRR 20-0400 08/18/20 B 3.3.1-18 84 DRR 20-0400 08/18/20 B 3.3.1-19 84 DRR 20-0400 08/18/20 B 3.3.1-20 84 DRR 20-0400 08/18/20 B 3.3.1-21 84 DRR 20-0400 08/18/20 B 3.3.1-22 84 DRR 20-0400 08/18/20 B 3.3.1-23 84 DRR 20-0400 08/18/20 B 3.3.1-24 84 DRR 20-0400 08/18/20 B 3.3.1-25 84 DRR 20-0400 08/18/20 B 3.3.1-26 84 DRR 20-0400 08/18/20 B 3.3.1-27 84 DRR 20-0400 08/18/20 B 3.3.1-28 84 DRR 20-0400 08/18/20 B 3.3.1-29 84 DRR 20-0400 08/18/20 B 3.3.1-30 84 DRR 20-0400 08/18/20 B 3.3.1-31 84 DRR 20-0400 08/18/20 B 3.3.1-32 84 DRR 20-0400 08/18/20 B 3.3.1-33 84 DRR 20-0400 08/18/20 Wolf Creek - Unit 1 iii Revision 89

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.1-34 84 DRR 20-0400 08/18/20 B 3.3.1-35 84 DRR 20-0400 08/18/20 B 3.3.1-36 84 DRR 20-0400 08/18/20 B 3.3.1-37 84 DRR 20-0400 08/18/20 B 3.3.1-38 84 DRR 20-0400 08/18/20 B 3.3.1-39 84 DRR 20-0400 08/18/20 B 3.3.1-40 84 DRR 20-0400 08/18/20 B 3.3.1-41 84 DRR 20-0400 08/18/20 B 3.3.1-42 84 DRR 20-0400 08/18/20 B 3.3.1-43 84 DRR 20-0400 08/18/20 B 3.3.1-44 84 DRR 20-0400 08/18/20 B 3.3.1-45 84 DRR 20-0400 08/18/20 B 3.3.1-46 84 DRR 20-0400 08/18/20 B 3.3.1-47 89 DRR 21-0966 7/7/21 B 3.3.1-48 89 DRR 21-0966 7/7/21 B 3.3.1-49 89 DRR 21-0966 7/7/21 B 3.3.1-50 89 DRR 21-0966 7/7/21 B 3.3.1-51 89 DRR 21-0966 7/7/21 B 3.3.1-52 89 DRR 21-0966 7/7/21 B 3.3.1-53 89 DRR 21-0966 7/7/21 B 3.3.1-54 89 DRR 21-0966 7/7/21 B 3.3.1-55 89 DRR 21-0966 7/7/21 B 3.3.1-56 89 DRR 21-0966 7/7/21 B 3.3.1-57 89 DRR 21-0966 7/7/21 B 3.3.1-58 89 DRR 21-0966 7/7/21 B 3.3.1-59 89 DRR 21-0966 7/7/21 B 3.3.1-60 89 DRR 21-0966 7/7/21 B 3.3.1-61 89 DRR 21-0966 7/7/21 B 3.3.1-62 (Del) 89 DRR 21-0966 7/7/21 B 3.3.2-1 84 DRR 20-0400 08/18/20 B 3.3.2-2 0 Amend. No. 123 12/18/99 B 3.3.2-3 0 Amend. No. 123 12/18/99 B 3.3.2-4 0 Amend. No. 123 12/18/99 B 3.3.2-5 0 Amend. No. 123 12/18/99 B 3.3.2-6 7 DRR 01-0474 5/1/01 B 3.3.2-7 0 Amend. No. 123 12/18/99 B 3.3.2-8 0 Amend. No. 123 12/18/99 B 3.3.2-9 0 Amend. No. 123 12/18/99 B 3.3.2-10 0 Amend. No. 123 12/18/99 B 3.3.2-11 0 Amend. No. 123 12/18/99 B 3.3.2-12 81 DRR 19-1027 10/28/19 B 3.3.2-13 0 Amend. No. 123 12/18/99 B 3.3.2-14 2 DRR 00-0147 4/24/00 B 3.3.2-15 0 Amend. No. 123 12/18/99 B 3.3.2-16 0 Amend. No. 123 12/18/99 B 3.3.2-17 0 Amend. No. 123 12/18/99 B 3.3.2-18 0 Amend. No. 123 12/18/99 B 3.3.2-19 37 DRR 08-0503 4/8/08 B 3.3.2-20 37 DRR 08-0503 4/8/08 B 3.3.2-21 37 DRR 08-0503 4/8/08 B 3.3.2-22 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 iv Revision 89

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.2-23 37 DRR 08-0503 4/8/08 B 3.3.2-24 39 DRR 08-1096 8/28/08 B 3.3.2-25 39 DRR 08-1096 8/28/08 B 3.3.2-26 39 DRR 08-1096 8/28/08 B 3.3.2-27 37 DRR 08-0503 4/8/08 B 3.3.2-28 84 DRR 20-0400 08/18/20 B 3.3.2-29 0 Amend. No. 123 12/18/99 B 3.3.2-30 0 Amend. No. 123 12/18/99 B 3.3.2-31 52 DRR 11-0724 4/11/11 B 3.3.2-32 52 DRR 11-0724 4/11/11 B 3.3.2-33 0 Amend. No. 123 12/18/99 B 3.3.2-34 0 Amend. No. 123 12/18/99 B 3.3.2-35 20 DRR 04-1533 2/16/05 B 3.3.2-36 20 DRR 04-1533 2/16/05 B 3.3.2-37 20 DRR 04-1533 2/16/05 B 3.3.2-38 20 DRR 04-1533 2/16/05 B 3.3.2-39 25 DRR 06-0800 5/18/06 B 3.3.2-40 20 DRR 04-1533 2/16/05 B 3.3.2-41 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-42 45 Amend. No. 187 (ETS) 3/5/10 B 3.3.2-43 20 DRR 04-1533 2/16/05 B 3.3.2-44 20 DRR 04-1533 2/16/05 B 3.3.2-45 89 DRR 21-0966 7/7/21 B 3.3.2-46 89 DRR 21-0966 7/7/21 B 3.3.2-47 89 DRR 21-0966 7/7/21 B 3.3.2-48 89 DRR 21-0966 7/7/21 B 3.3.2-49 89 DRR 21-0966 7/7/21 B 3.3.2-50 89 DRR 21-0966 7/7/21 B 3.3.2-51 89 DRR 21-0966 7/7/21 B 3.3.2-52 89 DRR 21-0966 7/7/21 B 3.3.2-53 43 DRR 09-1416 9/2/09 B 3.3.2-54 43 DRR 09-1416 9/2/09 B 3.3.2-55 43 DRR 09-1416 9/2/09 B 3.3.2-56 43 DRR 09-1416 9/2/09 B 3.3.2-57 43 DRR 09-1416 9/2/09 B 3.3.3-1 0 Amend. No. 123 12/18/99 B 3.3.3-2 5 DRR 00-1427 10/12/00 B 3.3.3-3 0 Amend. No. 123 12/18/99 B 3.3.3-4 0 Amend. No. 123 12/18/99 B 3.3.3-5 0 Amend. No. 123 12/18/99 B 3.3.3-6 8 DRR 01-1235 9/19/01 B 3.3.3-7 21 DRR 05-0707 4/20/05 B 3.3.3-8 81 DRR 19-1027 10/28/19 B 3.3.3-9 8 DRR 01-1235 9/19/01 B 3.3.3-10 19 DRR 04-1414 10/12/04 B 3.3.3-11 19 DRR 04-1414 10/12/04 B 3.3.3-12 21 DRR 05-0707 4/20/05 B 3.3.3-13 89 DRR 21-0966 7/7/21 B 3.3.3-14 89 DRR 21-0966 7/7/21 B 3.3.3-15 8 DRR 01-1235 9/19/01 B 3.3.4-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 v Revision 89

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TAB - B 3.3 INSTRUMENTATION (continued)

B 3.3.4-2 9 DRR 02-1023 2/28/02 B 3.3.4-3 15 DRR 03-0860 7/10/03 B 3.3.4-4 19 DRR 04-1414 10/12/04 B 3.3.4-5 89 DRR 21-0966 7/7/21 B 3.3.4-6 89 DRR 21-0966 7/7/21 B 3.3.5-1 88 DRR 21-0591 4/28/21 B 3.3.5-2 88 DRR 21-0591 4/28/21 B 3.3.5-3 1 DRR 99-1624 12/18/99 B 3.3.5-4 1 DRR 99-1624 12/18/99 B 3.3.5-5 0 Amend. No. 123 12/18/99 B 3.3.5-6 89 DRR 21-0966 7/7/21 B 3.3.5-7 89 DRR 21-0966 7/7/21 B 3.3.6-1 81 DRR 19-1027 10/28/19 B 3.3.6-2 81 DRR 19-1027 10/28/19 B 3.3.6-3 0 Amend. No. 123 12/18/99 B 3.3.6-4 0 Amend. No. 123 12/18/99 B 3.3.6-5 89 DRR 21-0966 7/7/21 B 3.3.6-6 89 DRR 21-0966 7/7/21 B 3.3.6-7 89 DRR 21-0966 7/7/21 B 3.3.7-1 81 DRR 19-1027 10/28/19 B 3.3.7-2 81 DRR 19-1027 10/28/19 B 3.3.7-3 57 DRR 13-0006 1/16/13 B 3.3.7-4 0 Amend. No. 123 12/18/99 B 3.3.7-5 0 Amend. No. 123 12/18/99 B 3.3.7-6 89 DRR 21-0966 7/7/21 B 3.3.7-7 89 DRR 21-0966 7/7/21 B 3.3.7-8 89 DRR 21-0966 7/7/21 B 3.3.8-1 84 DRR 20-0400 8/18/20 B 3.3.8-2 0 Amend. No. 123 12/18/99 B 3.3.8-3 57 DRR 13-0006 1/16/13 B 3.3.8-4 57 DRR 13-0006 1/16/13 B 3.3.8-5 89 DRR 21-0966 7/7/21 B 3.3.8-6 89 DRR 21-0966 7/7/21 B 3.3.8-7 89 DRR 21-0966 7/7/21 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1-1 84 DRR 20-0400 08/18/20 B 3.4.1-2 84 DRR 20-0400 08/18/20 B 3.4.1-3 10 DRR 02-0411 4/5/02 B 3.4.1-4 0 Amend. No. 123 12/18/99 B 3.4.1-5 89 DRR 21-0966 7/7/21 B 3.4.1-6 89 DRR 21-0966 7/7/21 B 3.4.2-1 0 Amend. No. 123 12/18/99 B 3.4.2-2 0 Amend. No. 123 12/18/99 B 3.4.2-3 89 DRR 21-0966 7/7/21 B 3.4.3-1 67 DRR 15-0116 2/10/15 B 3.4.3-2 0 Amend. No. 123 12/18/99 B 3.4.3-3 0 Amend. No. 123 12/18/99 B 3.4.3-4 0 Amend. No. 123 12/18/99 B 3.4.3-5 0 Amend. No. 123 12/18/99 B 3.4.3-6 89 DRR 21-0966 7/7/21 Wolf Creek - Unit 1 vi Revision 89

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.3-7 0 Amend. No. 123 12/18/99 B 3.4.4-1 0 Amend. No. 123 12/18/99 B 3.4.4-2 29 DRR 06-1984 10/17/06 B 3.4.4-3 89 DRR 21-0966 7/7/21 B 3.4.5-1 0 Amend. No. 123 12/18/99 B 3.4.5-2 53 DRR 11-1513 7/18/11 B 3.4.5-3 29 DRR 06-1984 10/17/06 B 3.4.5-4 0 Amend. No. 123 12/18/99 B 3.4.5-5 29 DRR 21-0966 7/7/21 B 3.4.5-6 89 DRR 21-0966 7/7/21 B 3.4.6-1 53 DRR 11-1513 7/18/11 B 3.4.6-2 72 DRR 15-1918 10/26/15 B 3.4.6-3 12 DRR 02-1062 9/26/02 B 3.4.6-4 89 DRR 21-0966 7/7/21 B 3.4.6-5 75 DRR 16-1909 10/26/16 B 3.4.6-6 89 DRR 21-0966 7/7/21 B 3.4.7-1 12 DRR 02-1062 9/26/02 B 3.4.7-2 17 DRR 04-0453 5/26/04 B 3.4.7-3 72 DRR 15-1918 10/26/15 B 3.4.7-4 89 DRR 21-0966 7/7/21 B 3.4.7-5 89 DRR 21-0966 7/7/21 B 3.4.7-6 89 DRR 21-0966 7/7/21 B 3.4.8-1 53 DRR 11-1513 7/18/11 B 3.4.8-2 72 DRR 15-1918 10/26/15 B 3.4.8-3 89 DRR 21-0966 7/7/21 B 3.4.8-4 89 DRR 21-0966 7/7/21 B 3.4.8-5 89 DRR 21-0966 7/7/21 B 3.4.9-1 0 Amend. No. 123 12/18/99 B 3.4.9-2 0 Amend. No. 123 12/18/99 B 3.4.9-3 0 Amend. No. 123 12/18/99 B 3.4.9-4 89 DRR 21-0966 7/7/21 B 3.4.10-1 5 DRR 00-1427 10/12/00 B 3.4.10-2 5 DRR 00-1427 10/12/00 B 3.4.10-3 0 Amend. No. 123 12/18/99 B 3.4.10-4 32 DRR 07-0139 2/7/07 B 3.4.11-1 0 Amend. No. 123 12/18/99 B 3.4.11-2 1 DRR 99-1624 12/18/99 B 3.4.11-3 19 DRR 04-1414 10/12/04 B 3.4.11-4 0 Amend. No. 123 12/18/99 B 3.4.11-5 1 DRR 99-1624 12/18/99 B 3.4.11-6 0 Amend. No. 123 12/18/99 B 3.4.11-7 89 DRR 21-0966 7/7/21 B 3.4.12-1 61 DRR 14-0346 2/27/14 B 3.4.12-2 61 DRR 14-0346 2/27/14 B 3.4.12-3 0 Amend. No. 123 12/18/99 B 3.4.12-4 61 DRR 14-0346 2/27/14 B 3.4.12-5 61 DRR 14-0346 2/27/14 B 3.4.12-6 56 DRR 12-1792 11/7/12 B 3.4.12-7 61 DRR 14-0346 2/27/14 B 3.4.12-8 1 DRR 99-1624 12/18/99 B 3.4.12-9 56 DRR 12-1792 11/7/12 Wolf Creek - Unit 1 vii Revision 89

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TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.12-10 0 Amend. No. 123 12/18/99 B 3.4.12-11 61 DRR 14-0346 2/27/14 B 3.4.12-12 89 DRR 21-0966 7/7/21 B 3.4.12-13 89 DRR 21-0966 7/7/21 B 3.4.12-14 89 DRR 21-0966 7/7/21 B 3.4.13-1 0 Amend. No. 123 12/18/99 B 3.4.13-2 81 DRR 19-1027 10/28/19 B 3.4.13-3 29 DRR 06-1984 10/17/06 B 3.4.13-4 35 DRR 07-1553 9/28/07 B 3.4.13-5 89 DRR 21-0966 7/7/21 B 3.4.13-6 89 DRR 21-0966 7/7/21 B 3.4.14-1 0 Amend. No. 123 12/18/99 B 3.4.14-2 0 Amend. No. 123 12/18/99 B 3.4.14-3 0 Amend. No. 123 12/18/99 B 3.4.14-4 0 Amend. No. 123 12/18/99 B 3.4.14-5 89 DRR 21-0966 7/7/21 B 3.4.14-6 89 DRR 21-0966 7/7/21 B 3.4.15-1 31 DRR 06-2494 12/13/06 B 3.4.15-2 31 DRR 06-2494 12/13/06 B 3.4.15-3 33 DRR 07-0656 5/1/07 B 3.4.15-4 33 DRR 07-0656 5/1/07 B 3.4.15-5 65 DRR 14-2146 9/30/14 B 3.4.15-6 31 DRR 06-2494 12/13/06 B 3.4.15-7 89 DRR 21-0966 7/7/21 B 3.4.15-8 31 DRR 06-2494 12/13/06 B 3.4.16-1 81 DRR 19-1027 10/28/19 B 3.4.16-2 84 DRR 20-0400 08/18/20 B 3.4.16-3 31 DRR 06-2494 12/13/06 B 3.4.16-4 89 DRR 21-0966 7/7/21 B 3.4.16-5 89 DRR 21-0966 7/7/21 B 3.4.17-1 29 DRR 06-1984 10/17/06 B 3.4.17-2 81 DRR 19-1027 10/28/19 B 3.4.17-3 52 DRR 11-0724 4/11/11 B 3.4.17-4 81 DRR 19-1027 10/28/19 B 3.4.17-5 57 DRR 13-0006 1/16/13 B 3.4.17-6 57 DRR 13-0006 1/16/13 B 3.4.17-7 81 DRR 19-1027 10/28/19 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1-1 0 Amend. No. 123 12/18/99 B 3.5.1-2 0 Amend. No. 123 12/18/99 B 3.5.1-3 73 DRR 15-2135 11/17/15 B 3.5.1-4 73 DRR 15-2135 11/17/15 B 3.5.1-5 1 DRR 99-1624 12/18/99 B 3.5.1-6 89 DRR 21-0966 7/7/21 B 3.5.1-7 89 DRR 21-0966 7/7/21 B 3.5.1-8 1 DRR 99-1624 12/18/99 B 3.5.2-1 84 DRR 20-0400 08/18/20 B 3.5.2-2 0 Amend. No. 123 12/18/99 B 3.5.2-3 0 Amend. No. 123 12/18/99 B 3.5.2-4 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 viii Revision 89

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IMPLEMENTED (4)

TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

B 3.5.2-5 72 DRR 15-1918 10/26/15 B 3.5.2-6 42 DRR 09-1009 7/16/09 B 3.5.2-7 89 DRR 21-0966 7/7/21 B 3.5.2-8 89 DRR 21-0966 7/7/21 B 3.5.2-9 89 DRR 21-0966 7/7/21 B 3.5.2-10 89 DRR 21-0966 7/7/21 B 3.5.2-11 89 DRR 21-0966 7/7/21 B 3.5.2-12 72 DRR 15-1918 10/26/15 B 3.5.3-1 56 DRR 12-1792 11/7/12 B 3.5.3-2 72 DRR 15-1918 10/26/15 B 3.5.3-3 56 DRR 12-1792 11/7/12 B 3.5.3-4 56 DRR 12-1792 11/7/12 B 3.5.4-1 0 Amend. No. 123 12/18/99 B 3.5.4-2 0 Amend. No. 123 12/18/99 B 3.5.4-3 0 Amend. No. 123 12/18/99 B 3.5.4-4 0 Amend. No. 123 12/18/99 B 3.5.4-5 89 DRR 21-0966 7/7/21 B 3.5.4-6 89 DRR 21-0966 7/7/21 B 3.5.5-1 21 DRR 05-0707 4/20/05 B 3.5.5-2 21 DRR 05-0707 4/20/05 B 3.5.5-3 89 DRR 21-0966 7/7/21 B 3.5.5-4 21 DRR 05-0707 4/20/05 TAB - B 3.6 CONTAINMENT SYSTEMS B 3.6.1-1 0 Amend. No. 123 12/18/99 B 3.6.1-2 81 DRR 19-1027 10/28/19 B 3.6.1-3 0 Amend. No. 123 12/18/99 B 3.6.1-4 87 DRR 21-0359 3/25/21 B 3.6.2-1 81 DRR 19-1027 10/28/19 B 3.6.2-2 0 Amend. No. 123 12/18/99 B 3.6.2-3 0 Amend. No. 123 12/18/99 B 3.6.2-4 0 Amend. No. 123 12/18/99 B 3.6.2-5 0 Amend. No. 123 12/18/99 B 3.6.2-6 89 DRR 21-0966 7/7/21 B 3.6.2-7 89 DRR 21-0966 7/7/21 B 3.6.3-1 0 Amend. No. 123 12/18/99 B 3.6.3-2 84 DRR 20-0400 08/18/20 B 3.6.3-3 81 DRR 19-1027 10/28/19 B 3.6.3-4 49 DRR 11-0014 1/31/11 B 3.6.3-5 49 DRR 11-0014 1/31/11 B 3.6.3-6 49 DRR 11-0014 1/31/11 B 3.6.3-7 41 DRR 09-0288 3/20/09 B 3.6.3-8 36 DRR 08-0255 3/11/08 B 3.6.3-9 41 DRR 21-0966 7/7/21 B 3.6.3-10 89 DRR 21-0966 7/7/21 B 3.6.3-11 36 DRR 08-0255 3/11/08 B 3.6.3-12 89 DRR 21-0966 7/7/21 B 3.6.3-13 89 DRR 21-0966 7/7/21 B 3.6.3-14 36 DRR 08-0255 3/11/08 B 3.6.3-15 39 DRR 08-1096 8/28/08 B 3.6.3-16 39 DRR 08-1096 8/28/08 Wolf Creek - Unit 1 ix Revision 89

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TAB - B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.3-17 36 DRR 08-0255 3/11/08 B 3.6.3-18 36 DRR 08-0255 3/11/08 B 3.6.3-19 36 DRR 08-0255 3/11/08 B 3.6.4-1 39 DRR 08-1096 8/28/08 B 3.6.4-2 0 Amend. No. 123 12/18/99 B 3.6.4-3 89 DRR 21-0966 7/7/21 B 3.6.5-1 0 Amend. No. 123 12/18/99 B 3.6.5-2 37 DRR 08-0503 4/8/08 B 3.6.5-3 89 DRR 21-0966 7/7/21 B 3.6.5-4 0 Amend. No. 123 12/18/99 B 3.6.6-1 81 DRR 19-1027 10/28/19 B 3.6.6-2 63 DRR 14-1572 7/1/14 B 3.6.6-3 37 DRR 08-0503 4/8/08 B 3.6.6-4 81 DRR 19-1027 10/28/19 B 3.6.6-5 0 Amend. No. 123 12/18/99 B 3.6.6-6 18 DRR 04-1018 9/1/04 B 3.6.6-7 89 DRR 21-0966 7/7/21 B 3.6.6-8 89 DRR 21-0966 7/7/21 B 3.6.6-9 72 DRR 15-1918 10/26/15 B 3.6.6-10 89 DRR 21-0966 7/7/21 B 3.6.6.11 80 DRR 19-0524 5/30/19 B 3.6.7-1 0 Amend. No. 123 12/18/99 B 3.6.7-2 81 DRR 19-1027 10/28/19 B 3.6.7-3 89 DRR 21-0966 7/7/21 B 3.6.7-4 89 DRR 21-0966 7/7/21 B 3.6.7-5 89 DRR 21-0966 7/7/21 TAB - B 3.7 PLANT SYSTEMS B 3.7.1-1 0 Amend. No. 123 12/18/99 B 3.7.1-2 84 DRR 20-0400 08/18/20 B 3.7.1-3 0 Amend. No. 123 12/18/99 B 3.7.1-4 84 DRR 20-0400 08/18/20 B 3.7.1-5 84 DRR 20-0400 08/18/20 B 3.7.1-6 84 DRR 20-0400 08/18/20 B 3.7.2-1 44 DRR 09-1744 10/28/09 B 3.7.2-2 44 DRR 09-1744 10/28/09 B 3.7.2-3 44 DRR 09-1744 10/28/09 B 3.7.2-4 81 DRR 19-1027 10/28/19 B 3.7.2-5 44 DRR 09-1744 10/28/09 B 3.7.2-6 44 DRR 09-1744 10/28/09 B 3.7.2-7 44 DRR 09-1744 10/28/09 B 3.7.2-8 44 DRR 09-1744 10/28/09 B 3.7.2-9 89 DRR 21-0966 7/7/21 B 3.7.2-10 81 DRR 19-1027 10/28/19 B 3.7.2-11 44 DRR 09-1744 10/28/09 B 3.7.3-1 37 DRR 08-0503 4/8/08 B 3.7.3-2 50 DRR 11-0449 3/9/11 B 3.7.3-3 37 DRR 08-0503 4/8/08 B 3.7.3-4 37 DRR 08-0503 4/8/08 B 3.7.3-5 37 DRR 08-0503 4/8/08 B 3.7.3-6 37 DRR 08-0503 4/8/08 Wolf Creek - Unit 1 x Revision 89

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.3-7 37 DRR 08-0503 4/8/08 B 3.7.3-8 37 DRR 08-0503 4/8/08 B 3.7.3-9 66 DRR 14-2329 11/6/14 B 3.7.3-10 89 DRR 21-0966 7/7/21 B 3.7.3-11 37 DRR 08-0503 4/8/08 B 3.7.4-1 1 DRR 99-1624 12/18/99 B 3.7.4-2 84 DRR 20-0400 08/18/20 B 3.7.4-3 19 DRR 04-1414 10/12/04 B 3.7.4-4 19 DRR 04-1414 10/12/04 B 3.7.4-5 89 DRR 21-0966 7/7/21 B 3.7.5-1 54 DRR 11-2394 11/16/11 B 3.7.5-2 54 DRR 11-2394 11/16/11 B 3.7.5-3 0 Amend. No. 123 12/18/99 B 3.7.5-4 85 DRR 20-0988 10/24/20 B 3.7.5-5 76 DRR 17-0343 2/21/17 B 3.7.5-6 85 DRR 20-0988 10/24/20 B 3.7.5-7 89 DRR 21-0966 7/7/21 B 3.7.5-8 89 DRR 21-0966 7/7/21 B 3.7.5-9 89 DRR 21-0966 7/7/21 B 3.7.5-10 85 DRR 20-0988 10/24/20 B 3.7.6-1 0 Amend. No. 123 12/18/99 B 3.7.6-2 0 Amend. No. 123 12/18/99 B 3.7.6-3 89 DRR 21-0966 7/7/21 B 3.7.7-1 0 Amend. No. 123 12/18/99 B 3.7.7-2 0 Amend. No. 123 12/18/99 B 3.7.7-3 89 DRR 21-0966 7/7/21 B 3.7.7-4 89 DRR 21-0966 7/7/21 B 3.7.8-1 0 Amend. No. 123 12/18/99 B 3.7.8-2 0 Amend. No. 123 12/18/99 B 3.7.8-3 0 Amend. No. 123 12/18/99 B 3.7.8-4 89 DRR 21-0966 7/7/21 B 3.7.8-5 89 DRR 21-0966 7/7/21 B 3.7.9-1 3 Amend. No. 134 7/14/00 B 3.7.9-2 3 Amend. No. 134 7/14/00 B 3.7.9-3 89 DRR 21-0966 7/7/21 B 3.7.9-4 3 Amend. No. 134 7/14/00 B 3.7.10-1 64 DRR 14-1822 8/28/14 B 3.7.10-2 81 DRR 19-1027 10/28/19 B 3.7.10-3 81 DRR 19-1027 10/28/19 B 3.7.10-4 81 DRR 19-1027 10/28/19 B 3.7.10-5 81 DRR 19-1027 10/28/19 B 3.7.10-6 57 DRR 13-0006 1/16/13 B 3.7.10-7 89 DRR 21-0966 7/7/21 B 3.7.10-8 89 DRR 21-0966 7/7/21 B 3.7.10-9 81 DRR 19-1027 10/28/19 B 3.7.11-1 0 Amend. No. 123 12/18/99 B 3.7.11-2 57 DRR 13-0006 1/16/13 B 3.7.11-3 89 DRR 21-0966 7/7/21 B 3.7.11-4 63 DRR 14-1572 7/1/14 B 3.7.12-1 0 Amend. No. 123 12/18/99 B 3.7.13-1 24 DRR 06-0051 2/28/06 Wolf Creek - Unit 1 xi Revision 89

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TAB - B 3.7 PLANT SYSTEMS (continued)

B 3.7.13-2 81 DRR 19-1027 10/28/19 B 3.7.13-3 81 DRR 19-1027 10/28/19 B 3.7.13-4 81 DRR 19-1027 10/28/19 B 3.7.13-5 89 DRR 21-0966 7/7/21 B 3.7.13-6 89 DRR 21-0966 7/7/21 B 3.7.13-7 89 DRR 21-0966 7/7/21 B 3.7.13-8 89 DRR 21-0966 7/7/21 B 3.7.14-1 0 Amend. No. 123 12/18/99 B 3.7.15-1 81 DRR 19-1027 10/28/19 B 3.7.15-2 89 DRR 21-0966 7/7/21 B 3.7.15-3 81 DRR 19-1027 10/28/19 B 3.7.16-1 5 DRR 00-1427 10/12/00 B 3.7.16-2 23 DRR 05-1995 9/28/05 B 3.7.16-3 89 DRR 21-0966 7/7/21 B 3.7.17-1 7 DRR 01-0474 5/1/01 B 3.7.17-2 7 DRR 01-0474 5/1/01 B 3.7.17-3 5 DRR 00-1427 10/12/00 B 3.7.18-1 81 DRR 19-1027 10/28/19 B 3.7.18-2 81 DRR 19-1027 10/28/19 B 3.7.18-3 89 DRR 21-0966 7/7/21 B 3.7.19-1 44 DRR 09-1744 10/28/09 B 3.7.19-2 54 DRR 11-2394 11/16/11 B 3.7.19-3 54 DRR 11-2394 11/16/11 B 3.7.19-4 61 DRR 14-0346 2/27/14 B 3.7.19-5 61 DRR 14-0346 2/27/14 B 3.7.19-6 89 DRR 21-0966 7/7/21 B 3.7.19-7 89 DRR 21-0966 7/7/21 B 3.7.20-1 79 DRR 18-1579 10/22/18 B 3.7.20-2 79 DRR 18-1579 10/22/18 B 3.7.20-3 85 DRR 20-0988 10/24/20 B 3.7.20-4 89 DRR 21-0966 7/7/21 B 3.7.20-5 89 DRR 21-0966 7/7/21 TAB - B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1-1 88 DRR 21-0591 4/28/21 B 3.8.1-2 88 DRR 21-0591 4/28/21 B 3.8.1-3 88 DRR 21-0951 4/28/21 B 3.8.1-4 86 DRR 21-0212 2/18/21 B 3.8.1-5 59 DRR 13-1524 6/26/13 B 3.8.1-6 25 DRR 06-0800 5/18/06 B 3.8.1-7 26 DRR 06-1350 7/24/06 B 3.8.1-8 35 DRR 07-1553 9/28/07 B 3.8.1-9 42 DRR 09-1009 7/16/09 B 3.8.1-10 39 DRR 08-1096 8/28/08 B 3.8.1-11 36 DRR 08-0255 3/11/08 B 3.8.1-12 75 DRR 16-1909 10/26/16 B 3.8.1-13 47 DRR 10-1089 6/16/10 B 3.8.1-14 47 DRR 10-1089 6/16/10 B 3.8.1-15 47 DRR 10-1089 6/16/10 B 3.8.1-16 26 DRR 06-1350 7/24/06 B 3.8.1-17 26 DRR 06-1350 7/24/06 Wolf Creek - Unit 1 xii Revision 89

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.1-18 59 DRR 13-1524 6/26/13 B 3.8.1-19 89 DRR 21-0966 7/7/21 B 3.8.1-20 89 DRR 21-0966 7/7/21 B 3.8.1-21 89 DRR 21-0966 7/7/21 B 3.8.1-22 89 DRR 21-0966 7/7/21 B 3.8.1-23 89 DRR 16-1182 7/7/16 B 3.8.1-24 74 DRR 16-1182 7/7/16 B 3.8.1-25 89 DRR 21-0966 7/7/21 B 3.8.1-26 89 DRR 21-0966 7/7/21 B 3.8.1-27 89 DRR 21-0966 7/7/21 B 3.8.1-28 89 DRR 21-0966 7/7/21 B 3.8.1-29 89 DRR 21-0966 7/7/21 B 3.8.1-30 89 DRR 21-0966 7/7/21 B 3.8.1-31 89 DRR 21-0966 7/7/21 B 3.8.1-32 89 DRR 21-0966 7/7/21 B 3.8.1-33 89 DRR 21-0966 7/7/21 B 3.8.1-34 74 DRR 16-1182 7/7/16 B 3.8.2-1 57 DRR 13-0006 1/16/13 B 3.8.2-2 0 Amend. No. 123 12/18/99 B 3.8.2-3 86 DRR 21-0212 2/18/21 B 3.8.2-4 88 DRR 21-0591 4/28/21 B 3.8.2-5 57 DRR 13-0006 1/16/13 B 3.8.2-6 57 DRR 13-0006 1/16/13 B 3.8.2-7 57 DRR 13-0006 1/16/13 B 3.8.3-1 1 DRR 99-1624 12/18/99 B 3.8.3-2 0 Amend. No. 123 12/18/99 B 3.8.3-3 0 Amend. No. 123 12/18/99 B 3.8.3-4 1 DRR 99-1624 12/18/99 B 3.8.3-5 0 Amend. No. 123 12/18/99 B 3.8.3-6 89 DRR 21-0966 7/7/21 B 3.8.3-7 12 DRR 02-1062 9/26/02 B 3.8.3-8 89 DRR 21-0966 7/7/21 B 3.8.3-9 0 Amend. No. 123 12/18/99 B 3.8.4-1 0 Amend. No. 123 12/18/99 B 3.8.4-2 0 Amend. No. 123 12/18/99 B 3.8.4-3 0 Amend. No. 123 12/18/99 B 3.8.4-4 0 Amend. No. 123 12/18/99 B 3.8.4-5 89 DRR 21-0966 7/7/21 B 3.8.4-6 89 DRR 21-0966 7/7/21 B 3.8.4-7 6 DRR 00-1541 3/13/01 B 3.8.4-8 89 DRR 21-0966 7/7/21 B 3.8.4-9 89 DRR 21-0966 7/7/21 B 3.8.5-1 57 DRR 13-0006 1/16/13 B 3.8.5-2 0 Amend. No. 123 12/18/99 B 3.8.5-3 57 DRR 13-0006 1/16/13 B 3.8.5-4 57 DRR 13-0006 1/16/13 B 3.8.5-5 57 DRR 13-0006 1/16/13 B 3.8.6-1 0 Amend. No. 123 12/18/99 B 3.8.6-2 0 Amend. No. 123 12/18/99 B 3.8.6-3 89 DRR 21-0966 7/7/21 B 3.8.6-4 89 DRR 21-0966 7/7/21 Wolf Creek - Unit 1 xiii Revision 89

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TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.6-5 0 Amend. No. 123 12/18/99 B 3.8.6-6 0 Amend. No. 123 12/18/99 B 3.8.7-1 69 DRR 15-0493 3/26/15 B 3.8.7-2 69 DRR 15-0493 3/26/15 B 3.8.7-3 89 DRR 21-0966 7/7/21 B 3.8.7-4 0 Amend. No. 123 12/18/99 B 3.8.8-1 57 DRR 13-0006 1/16/13 B 3.8.8-2 0 Amend. No. 123 12/18/99 B 3.8.8-3 69 DRR 15-0493 3/26/15 B 3.8.8-4 57 DRR 13-0006 1/16/13 B 3.8.8-5 69 DRR 15-0493 3/26/15 B 3.8.9-1 54 DRR 11-2394 11/16/11 B 3.8.9-2 69 DRR 15-0493 3/26/15 B 3.8.9-3 54 DRR 11-2394 11/16/11 B 3.8.9-4 0 Amend. No. 123 12/18/99 B 3.8.9-5 69 DRR 15-0493 3/26/15 B 3.8.9-6 0 Amend. No. 123 12/18/99 B 3.8.9-7 0 Amend. No. 123 12/18/99 B 3.8.9-8 89 DRR 21-0966 7/7/21 B 3.8.9-9 0 Amend. No. 123 12/18/99 B 3.8.10-1 57 DRR 13-0006 1/16/13 B 3.8.10-2 0 Amend. No. 123 12/18/99 B 3.8.10-3 0 Amend. No. 123 12/18/99 B 3.8.10-4 57 DRR 13-0006 1/16/13 B 3.8.10-5 57 DRR 13-0006 1/16/13 B 3.8.10-6 89 DRR 21-0966 7/7/21 TAB - B 3.9 REFUELING OPERATIONS B 3.9.1-1 0 Amend. No. 123 12/18/99 B 3.9.1-2 19 DRR 04-1414 10/12/04 B 3.9.1-3 19 DRR 04-1414 10/12/04 B 3.9.1-4 89 DRR 21-0966 7/7/21 B 3.9.2-1 0 Amend. No. 123 12/18/99 B 3.9.2-2 0 Amend. No. 123 12/18/99 B 3.9.2-3 89 DRR 21-0966 7/7/21 B 3.9.3-1 68 DRR 15-0248 2/26/15 B 3.9.3-2 68 DRR 15-0248 2/26/15 B 3.9.3-3 89 DRR 21-0966 7/7/21 B 3.9.3-4 89 DRR 21-0966 7/7/21 B 3.9.4-1 81 DRR 19-1027 10/28/19 B 3.9.4-2 13 DRR 02-1458 12/03/02 B 3.9.4-3 81 DRR 19-1027 10/28/19 B 3.9.4-4 23 DRR 05-1995 9/28/05 B 3.9.4-5 89 DRR 21-0966 7/7/21 B 3.9.4-6 89 DRR 21-0966 7/7/21 B 3.9.5-1 0 Amend. No. 123 12/18/99 B 3.9.5-2 72 DRR 15-1918 10/26/15 B 3.9.5-3 32 DRR 07-0139 2/7/07 B 3.9.5-4 89 DRR 21-0966 7/7/21 B 3.9.5-5 89 DRR 21-0966 7/7/21 B 3.9.6-1 0 Amend. No. 123 12/18/99 Wolf Creek - Unit 1 xiv Revision 89

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TAB - B 3.9 REFUELING OPERATIONS (continued)

B 3.9.6-2 72 DRR 15-1918 10/26/15 B 3.9.6-3 42 DRR 09-1009 7/16/09 B 3.9.6-4 89 DRR 21-0966 7/7/21 B 3.9.6-5 89 DRR 21-0966 7/7/21 B 3.9.7-1 81 DRR 19-1027 10/28/19 B 3.9.7-2 89 DRR 21-0966 7/7/21 B 3.9.7-3 81 DRR 19-1027 10/28/19 Note 1 The page number is listed on the center of the bottom of each page.

Note 2 The revision number is listed in the lower right hand corner of each page. The Revision number will be page specific.

Note 3 The change document will be the document requesting the change. Amendment No.

123 issued the improved Technical Specifications and associated Bases which affected each page. The NRC has indicated that Bases changes will not be issued with License Amendments. Therefore, the change document should be a DRR number in accordance with AP 26A-002.

Note 4 The date effective or implemented is the date the Bases pages are issued by Document Control.

Wolf Creek - Unit 1 xv Revision 89

ENCLOSURE Page 74 of 264 SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.

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

a. RCS boron concentration;

b. Control and shutdown rod position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

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

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

2. USAR, Section 15.1.5.

3. USAR, Section 15.4.6.

4. 10 CFR 50.67.

Wolf Creek - Unit 1 B 3.1.1-5 Revision 89

ENCLOSURE Page 75 of 264 Core Reactivity B 3.1.2 BASES ACTIONS B.1 (continued)

If the core reactivity cannot be restored to within the 1% k/k limit, 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 3 is not met, then the boration required by LCO 3.1.1 Required Action A.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control and shutdown rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization (if necessary) of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

2. USAR, Chapter 15.

Wolf Creek - Unit 1 B 3.1.2-5 Revision 89

ENCLOSURE Page 76 of 264

ENCLOSURE Page 77 of 264

ENCLOSURE Page 78 of 264 Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS B.1 (continued)

If the shutdown banks cannot be restored to within their insertion limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the unit must be brought to a MODE 3 where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip.

This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

2. 10 CFR 50.46.

3. USAR, Section 4.3.1.5.

Wolf Creek - Unit 1 B 3.1.5-4 Revision 89

ENCLOSURE Page 79 of 264 Control Bank Insertion Limits B 3.1.6 BASES ACTIONS C.1 (continued) experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

SR 3.1.6.2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. The verification of compliance consists of an observation that the static rod positions of those control banks not fully withdrawn from the core are within the limits specified in the COLR. Bank sequence and overlap must also be maintained during rod movement, implicit within the LCO. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.1.6-5 Revision 89

ENCLOSURE Page 80 of 264 PHYSICS TEST Exceptions - MODE 2 B 3.1.8 BASES SURVEILLANCE SR 3.1.8.2 REQUIREMENTS (continued) Verification that the RCS lowest operating loop T avg is 541 F will ensure that the unit is not operating in a condition that could invalidate the safety analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.1.8.3 Verification that the THERMAL POWER is 5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.1.8.4 Verification that the SDM is within limits specified in the COLR ensures that, for the specific RCCA and RCS temperature manipulations performed during PHYSICS TESTS, the plant is not operating in a condition that could invalidate the safety analysis assumptions.

The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration;

b. Control bank position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

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

Wolf Creek - Unit 1 B 3.1.8-5 Revision 89

ENCLOSURE Page 81 of 264 PHYSICS TEST Exceptions - MODE 2 B 3.1.8 BASES SURVEILLANCE SR 3.1.8.4 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August, 1978.

4. WCAP-9272-P-A, Revision 0, Westinghouse Reload Safety Methodology, July 1985.

Wolf Creek - Unit 1 B 3.1.8-6 Revision 89

ENCLOSURE Page 82 of 264 RCS Boron Limitations < 500 F B 3.1.9 BASES SURVEILLANCE SR 3.1.9.1 REQUIREMENTS This SR ensures that the RCS boron concentration is within limit. The boron concentration is determined periodically by chemical analysis.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. Westinghouse Nuclear Safety Advisory Letter NSAL-00-016, Rod Withdrawal from Subcritical Protection in Lower Modes, December 4, 2000.

2. USAR, Section 15.4.1.

3. Technical Specification Bases Table B 3.3.1-1.

Wolf Creek - Unit 1 B 3.1.9-5 Revision 89

ENCLOSURE Page 83 of 264 FQ(Z) (FQ Methodology)

B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 (continued)

REQUIREMENTS If THERMAL POWER has been increased by 10% RTP since the last determination of FQC(Z), another evaluation of this factor is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions at this higher power level (to ensure that FQC(Z) values are being reduced sufficiently with power increase to stay within the LCO limits).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the F Q(Z) limits. Because power distribution measurements are taken at or near equilibrium conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the measurements. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation. The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, F QC(Z), by W(Z) gives the maximum F Q(Z) calculated to occur in normal operation, FQW(Z).

The limit with which FQW(Z) is compared varies inversely with power and directly with the function K(Z) provided in the COLR.

The W(Z) are provided for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations. F QW(Z) evaluations are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% inclusive; and

b. Upper core region, from 85 to 100% inclusive.

The amount of the axial core region that can be excluded during the performance of SR 3.2.1.2 shall not exceed 15% of the upper and lower core regions, and may be reduced on a cycle-specific basis as determined during the core reload design process. The amount of the axial core region that can be excluded during the performance of SR 3.2.1.2 is identified in the COLR. The axial core regions are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty Wolf Creek - Unit 1 B 3.2.1-8 Revision 89

ENCLOSURE Page 84 of 264 FQ(Z) (FQ Methodology)

B 3.2.1 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS of making a precise measurement in these regions. It should be noted that while the transient FQ(Z) limits are not measured in these axial core regions, the analytical transient FQ(Z) limits in these axial core regions are demonstrated to be satisfied during the core reload design process.

This Surveillance has been modified by a Note that may require more frequent surveillances be performed. When F QC(Z) is measured, an evaluation of the expression below is required to account for any increase to FQ(Z) that may occur and cause the F Q(Z) limit to be exceeded before the next required FQ(Z) evaluation.

If the two most recent FQ(Z) evaluations show an increase in the expression C

FQ (Z) maximum over z K(Z) it is required to meet the FQ(Z) limit with the last FQW(Z) increased by the appropriate factor specified in the COLR, or to evaluate F Q(Z) more frequently, each 7 EFPD. These alternative requirements prevent F Q(Z) from exceeding its limit for any significant period of time without detection.

Performing the Surveillance in MODE 1 prior to exceeding 75% RTP ensures that the FQ(Z) limit will be met when RTP is achieved, because peaking factors are generally decreased as power level is increased.

FQ(Z) is verified at power levels 10% RTP above the THERMAL POWER of its last verification, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions to ensure that FQ(Z) is within its limit at higher power levels.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.2.1-9 Revision 89

ENCLOSURE Page 85 of 264 FN H B 3.2.2 BASES ACTIONS B.1 (continued)

Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS SR 3.2.2.1 is modified by a Note. The Note applies during power ascensions following a plant shutdown (leaving MODE 1). The Note allows for power ascensions if the surveillances are not current. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution measurement can be obtained. Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions to perform the measurement.

The value of FN H is determined by using either the movable incore detector system or the Power Distribution Monitoring System to obtain a power distribution measurement. A calculation determines the maximum value of FN H from the measured power distribution. The measured value of FN H must be increased by 4% (if using the movable incore detector system) or increased by U H% (if using the Power Distribution Monitoring System, where U H is determined as described in Reference 4, with a minimum value of 4%) to account for measurement uncertainty before making comparisons to the FN H limit After each refueling, FN H must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that F N H limits are met at the beginning of each fuel cycle.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.4.8.

2. 10 CFR 50, Appendix A, GDC 26.

3. 10 CFR 50.46.

4. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994 (including Addendum 4, September 2012).

Wolf Creek - Unit 1 B 3.2.2-6 Revision 89

ENCLOSURE Page 86 of 264 AFD (RAOC Methodology)

B 3.2.3 BASES APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis.

For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES.

ACTIONS A.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to

< 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD is within its specified limits.

Indicated AFD is monitored for each OPERABLE NIS excore channel at the specified Frequency. Two or more OPERABLE excore channels outside the AFD limits would constitute failure of this surveillance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control and FQ Surveillance Technical Specification, February 1994.

2. USAR, Chapter 7.

Wolf Creek - Unit 1 B 3.2.3-3 Revision 89

ENCLOSURE Page 87 of 264 QPTR B 3.2.4 BASES ACTIONS A.7 (continued)

Required Action A.7 is modified by a Note that states that the peaking factor surveillances must be completed when the excore detectors have been normalized to restore QPTR to within limit (i.e., Required Action A.6). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limit.

B.1 If Required Actions A.1 through A.7 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.

SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1 to confirm the indication of the remaining three excore channels.

This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the input from one Power Range Neutron Flux channel is inoperable and the THERMAL POWER is > 75% RTP.

Wolf Creek - Unit 1 B 3.2.4-6 Revision 89

ENCLOSURE Page 88 of 264 QPTR B 3.2.4 BASES SURVEILLANCE SR 3.2.4.2 (continued)

REQUIREMENTS With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

For purposes of monitoring the QPTR when one power range channel is inoperable, either the movable incore detector system or the Power Distribution Monitoring System is used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore QPTR. If one of the symmetric thimbles is not available, then other pairs (triples) of symmetric thimbles can be monitored to gain information about the quadrant with the out-of-service thimble, provided the reference case is set up with the same thimble groupings. Therefore, incore monitoring of QPTR can be used to confirm that QPTR is within limits.

With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the power distribution may be compared against previous core power distribution measurements either using the symmetric thimbles as described above or a complete core power distribution measurement.

REFERENCES 1. 10 CFR 50.46.

2. USAR, Section 15.4.8.

3. 10 CFR 50, Appendix A, GDC 26.

Wolf Creek - Unit 1 B 3.2.4-7 Revision 89

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ENCLOSURE Page 93 of 264 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.5 REQUIREMENTS (continued) SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.

Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function, including operation of the P-7 permissive which is a logic function only. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.6 SR 3.3.1.6 is a calibration of the excore channels to the core power distribution, measured using either the movable incore detector system or the Power Distribution Monitoring System. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the core power distribution measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f( I) input to the Overtemperature T Function.

A Note modifies SR 3.3.1.6. The Note states that this Surveillance is not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after achieving equilibrium conditions with THERMAL POWER 75% RTP. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to obtain a core power distribution measurement. The SR is deferred until a scheduled testing plateau above 75% RTP is attained during a power ascension. During a typical power ascension, it is usually necessary to control the axial flux difference at lower power levels through control rod insertion. After equilibrium conditions are achieved at the specified power plateau, a core power distribution measurement must be taken and the required data collected. The data is typically analyzed and the appropriate excore calibrations completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after achieving equilibrium conditions. An additional time allowance of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided during which the effects of equipment failures may be remedied and any required re-testing may be performed.

The allowance of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascensions and associated testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use.

Wolf Creek - Unit 1 B 3.3.1-51 Revision 89

ENCLOSURE Page 94 of 264 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT.

A COT is performed on each required channel to ensure the channel will perform the intended Function.

Setpoints must be within the Allowable Values specified in Table 3.3.1-1.

SR 3.3.1.7 is modified by a Note that provides a 4-hour delay in the requirement to perform this Surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs are open and SR 3.3.1.7 is no longer required to be performed. If the unit is to be in MODE 3 with the RTBs closed for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> this Surveillance must be performed prior to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entry into MODE 3. Note 2 requires that the quarterly COT for the source range instrumentation shall include verification by observation of the associated permissive annunciator window that the P-6 and P-10 interlocks are in their required state for the existing conditions.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7, and it is modified by a Note that this test shall include verification that the P-6 and P-10 interlocks are in their required state for the existing unit conditions. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed, e.g., by observation of the associated permissive annunciator window, within the Frequency specified in the Surveillance Frequency Control Program of the Frequencies prior to reactor startup, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10, and four hours after reducing power below P-6. The Frequency of "prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a Wolf Creek - Unit 1 B 3.3.1-52 Revision 89

ENCLOSURE Page 95 of 264 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 (continued)

REQUIREMENTS delay to perform the testing required by this surveillance. The Frequency specified in the Surveillance Frequency Control Program applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reducing power below P-10, and four hours after reducing power below P-6. The MODE of Applicability for this surveillance is < P-10 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or < P-6 for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, then the testing required by this surveillance must be performed prior to the expiration of the 12-hour or the 4-hour limit. These time limits are reasonable, based on operating experience to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE (< P-10 or < P-6) for the periods discussed above. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.9 SR 3.3.1.9 is the performance of a TADOT. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that excludes verification of setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.

SR 3.3.1.10 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.1-53 Revision 89

ENCLOSURE Page 96 of 264 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.10 (continued)

REQUIREMENTS SR 3.3.1.10 is modified by a Note stating that this test shall include verification that the time constants are adjusted to the prescribed values where applicable. This does not include verification of time delay relays.

These are verified by response time testing per SR 3.3.1.16. Whenever an RTD is replaced in Functions 6 or 7, the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

SR 3.3.1.11 SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10. This SR is modified by three Notes. Note 1 states that neutron detectors are excluded from the CHANNEL CALIBRATION. The source range neutron detectors are maintained based on manufacturers recommendations. For the intermediate and power range channels, detector plateau curves are obtained, evaluated, and compared to manufacturers data. Note 2 states that this test shall include verification that the time constants are adjusted to the prescribed values where applicable. Note 3 states that the power and intermediate range detector plateau voltage verification is not required to be current until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after achieving equilibrium conditions with THERMAL POWER 95% RTP. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to perform a meaningful detector plateau voltage verification. The allowance of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascension testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The conditions for verifying the power and intermediate range detector plateau voltages are described above. The other remaining portions of the CHANNEL CALIBRATIONS may be performed either during a plant outage or during plant operation.

SR 3.3.1.12 Not Used.

Wolf Creek - Unit 1 B 3.3.1-54 Revision 89

ENCLOSURE Page 97 of 264 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.13 REQUIREMENTS (continued) SR 3.3.1.13 is the performance of a COT of RTS interlocks.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.14 SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip, the SI Input from ESFAS, and the Reactor Trip Bypass Breaker undervoltage trip mechanisms. The Manual Reactor Trip TADOT shall independently verify the OPERABILITY of the handswitch undervoltage and shunt trip contacts for both the Reactor Trip Breakers and Reactor Trip Bypass Breakers. The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip mechanism.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The SR is modified by a Note that excludes verification of setpoints from the TADOT. The Functions affected have no setpoints associated with them.

SR 3.3.1.15 SR 3.3.1.15 is the performance of a TADOT of Turbine Trip Functions.

This TADOT is as described in SR 3.3.1.4, except that this test is performed prior to exceeding the P-9 interlock whenever the unit has been in MODE 3. This Surveillance is not required if it has been performed within the previous 31 days. Verification of the Trip Setpoint does not have to be performed for this Surveillance. Performance of this test will ensure that the turbine trip Function is OPERABLE prior to exceeding the P-9 interlock.

SR 3.3.1.16 SR 3.3.1.16 verifies that the individual channel actuation response times are less than or equal to the maximum values assumed in the accident analysis. Response time testing acceptance criteria are included in Table B 3.3.1-2. No credit was taken in the safety analyses for those Wolf Creek - Unit 1 B 3.3.1-55 Revision 89

ENCLOSURE Page 98 of 264 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.16 (continued)

REQUIREMENTS channels with response times listed as N.A. No response time testing requirements apply where N.A. is listed in Table B 3.3.1-2. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor until loss of stationary gripper coil voltage.

For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time verification is performed with the time constants set at their nominal values. The response time may be measured by a series of overlapping tests, or other verification (e.g.,

Ref. 7), such that the entire response time is measured.

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated response times with actual response time tests on the remainder of the channel. Allocations for response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements (Ref. 7), provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

The allocations for sensor response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where response time could be affected is replacing the sensing assembly of a transmitter.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal.

Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input to the first electronic component in the channel.

Wolf Creek - Unit 1 B 3.3.1-56 Revision 89

ENCLOSURE Page 99 of 264 RTS Instrumentation B 3.3.1 BASES REFERENCES 1. USAR, Chapter 7.

2. USAR, Chapter 15.

3. IEEE-279-1971.

4. 10 CFR 50.49.

5. WCNOC Nuclear Safety Analysis Setpoint Methodology for the Reactor Protection System, (TR-89-0001).

6. WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996.

7. WCAP-9226, Reactor Core Response to Excessive Secondary Steam Releases, Revision 1, January 1978.

8. IE Information Notice 79-22, Qualification of Control Systems, September 14, 1979.

9. Wolf Creek Setpoint Methodology Report, SNP(KG)-492, August 29, 1984.

10. USAR, Table 15.0-4.

11. WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, October 1998.

12. WCAP-15376-P-A, Revision 1, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times, March 2003.

13. WOG-06-17, WCAP-10271-P-A Justification for Bypass Test Time and Completion Time Technical Specification Changes for Reactor Trip on Turbine Trip (ITSWG Action item #314), January 20, 2006.

Wolf Creek - Unit 1 B 3.3.1-57 Revision 89

ENCLOSURE Page 100 of 264 RTS Instrumentation B 3.3.1 TABLE B 3.3.1-1 (Page 1 of 2)

FUNCTION TRIP SETPOINT(a)

1. Manual Reactor Trip NA

2. Power Range Neutron Flux

a. High

b. Low

3. Power Range Neutron Flux

a. High Positive Rate

b. High Negative Rate

4. Intermediate Range Neutron Flux 5

5. Source Range Neutron Flux cps

6. See Table 3.3.1-1 Note 1

7. See Table 3.3.1-1 Note 2

8. Pressurizer Pressure

a. Low

b. High

9. Pressurizer Water level - High

10. Reactor Coolant Flow - Low Normalized Flow

11. Not Used

12. Undervoltage RCPs

13. Underfrequency RCPs 15 Hz

14. Steam Generator (SG) Water Level Low - Low instrument span

15. Not Used

16. Turbine Trip

a. Low Fluid Oil Pressure

b. Turbine Stop Valve Closure Wolf Creek - Unit 1 B 3.3.1-58 Revision 89

ENCLOSURE Page 101 of 264 RTS Instrumentation B 3.3.1 TABLE B 3.3.1-1 (Page 2 of 2)

FUNCTION TRIP SETPOINT(a)

17. Safety Injection (SI) Input from N.A.

Engineered Safety Feature Actuation System (ESFAS)

18. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6 -10 amps

b. Low Power Reactor Trips Block, P-7 N.A.

c. Power Range Neutron Flux, P-8

d. Power Range Neutron Flux, P-9

e. Power Range Neutron Flux, P-10 10% RTP

f. Turbine Impulse Pressure, P-13

19. Reactor Trip Breakers N.A.

20. Reactor Trip breaker Undervoltage and N.A.

Shunt Trip Mechanisms

21. Automatic Trip Logic N.A.

(a)

The inequality sign only indicates conservative direction. The as-left value will be within a two-sided calibration tolerance band on either side of the nominal value. This also applies to values.

Wolf Creek - Unit 1 B 3.3.1-59 Revision 89

ENCLOSURE Page 102 of 264 RTS Instrumentation B 3.3.1 TABLE B 3.3.1-2 (Page 1 of 2)

FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip N.A.

2. Power Range Neutron Flux (1)

a. High (1)

b. Low

3. Power Range Neutron Flux (1)

a. High Positive Rate .

(1)

b. High Negative Rate ond

4. Intermediate Range Neutron Flux N.A.

5. Source Range Neutron Flux N.A.

(1) 6.

(1) 7.

8. Pressurizer Pressure

a. Low

b. High 1.0 second

9. Pressurizer Water Level - High N.A.

10. Reactor Coolant Flow - Low

a. Single Loop (Above P-8)

b. Two Loops (Above P-7 and below P-8)

11. Not Used

12. Undervoltage - Reactor Coolant Pumps

13. Underfrequency - Reactor Coolant Pumps

14. Steam Generator Water Level - Low-Low

15. Not Used (1)

Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

Wolf Creek - Unit 1 B 3.3.1-60 Revision 89

ENCLOSURE Page 103 of 264 RTS Instrumentation B 3.3.1 TABLE B 3.3.1-2 (Page 2 of 2)

FUNCTIONAL UNIT RESPONSE TIME

16. Turbine Trip

a. Low Fluid Oil Pressure N.A.

b. Turbine Stop Valve Closure N.A.

17. Safety Injection Input for ESF N.A.

18. Reactor Trip System Interlocks N.A.

19. Reactor Trip Breakers N.A.

20. Reactor Trip Breaker Undervoltage and N.A.

Shunt Trip Mechanisms

21. Automatic Trip and Interlock Logic N.A.

Wolf Creek - Unit 1 B 3.3.1-61 Revision 89

ENCLOSURE Page 104 of 264 ESFAS Instrumentation B 3.3.2 BASES ACTIONS P.1, P.2.1, and P.2.2 (continued) redundancy provided by the motor driven AFW pumps, and the low probability of an event occurring during this interval. If the Function cannot be returned to OPERABLE status, the unit must be placed in MODE 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 4 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power in an orderly manner and without challenging unit systems. In MODE 4, the unit does not have any analyzed transients or conditions that require the turbine driven AFW pump for mitigation.

SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs column REQUIREMENTS of Table 3.3.2-1.

A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which SRs apply to which ESFAS Functions.

Note that each channel of process protection supplies both trains of the ESFAS. When testing channel I, train A and train B must be examined.

Similarly, train A and train B must be examined when testing channel II, channel III, and channel IV. The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

SR 3.3.2.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and reliability. If a channel is outside the criteria, it may be an indication Wolf Creek - Unit 1 B 3.3.2-45 Revision 89

ENCLOSURE Page 105 of 264 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.1 (continued)

REQUIREMENTS that the sensor or the signal processing equipment has drifted outside its limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.2.2 SR 3.3.2.2 is the performance of an ACTUATION LOGIC TEST using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. In addition, the master relay coil is pulse tested for continuity. This verifies that the logic modules are OPERABLE and that there is an intact voltage signal path to the master relay coils. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.2.3 SR 3.3.2.3 is the performance of an ACTUATION LOGIC TEST using the BOP ESFAS automatic tester. The continuity check does not have to be performed, as explained in the Note. This SR is applied to the balance of plant actuation logic and relays that do not have circuits installed to perform the continuity check. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. The time allowed for the testing (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) is justified in Reference 7. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.2-46 Revision 89

ENCLOSURE Page 106 of 264 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.5 REQUIREMENTS (continued) SR 3.3.2.5 is the performance of a COT.

A COT is performed on each required channel to ensure the channel will perform the intended Function. Setpoints must be found within the Allowable Values specified in Table 3.3.2-1.

The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.2.6 SR 3.3.2.6 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the slave relay blocking circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

For Function 4.c (Steam Line Isolation - Automatic Actuation Logic (MSFIS)) and Function 5.b (Turbine Trip and Feedwater Isolation -

Automatic Actuation Logic (MSFIS)), SR 3.3.2.6 is performed on the associated slave relays in the SSPS cabinets and includes verification that the slave relays are energized at the MSFIS cabinets.

Wolf Creek - Unit 1 B 3.3.2-47 Revision 89

ENCLOSURE Page 107 of 264 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.7 REQUIREMENTS (continued) SR 3.3.2.7 is the performance of a TADOT. This test is a check of the Loss of Offsite Power function. The trip actuating devices tested within the scope of SR 3.3.2.7 are the LSELS output relays and BOP ESFAS separation groups logic associated with the auto-start of the turbine driven AFW pump upon an ESF bus undervoltage condition.

The SR is modified by a Note that excludes verification of setpoints for relays. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The trip actuating devices tested have no associated setpoint.

SR 3.3.2.8 SR 3.3.2.8 is the performance of a TADOT. This test is a check of the Manual Actuation Functions (SSPS) and AFW pump start on trip of all MFW pumps BOP ESFAS. The Manual Safety Injection TADOT shall independently verify OPERABILITY of the handswitch undervoltage and shunt trip contacts for both the Reactor Trip Breakers and Reactor Trip Bypass Breakers as well as the contacts for safety injection actuation.

Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The SR is modified by a Note that excludes verification of setpoints during the TADOT for manual initiation Functions. The manual initiation Functions have no associated setpoints.

SR 3.3.2.9 SR 3.3.2.9 is the performance of a CHANNEL CALIBRATION.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.

Wolf Creek - Unit 1 B 3.3.2-48 Revision 89

ENCLOSURE Page 108 of 264 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.9 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note stating that this test shall include verification that the time constants are adjusted to the prescribed values where applicable. This does not include verification of time delay relays.

These are verified by response time testing per SR 3.3.2.10.

SR 3.3.2.10 This SR verifies the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.

Response Time verification acceptance criteria are included in Table B 3.3.2-2. Table B 3.3.2-2 format is based on the initiating trip signal. No credit was taken in the safety analyses for those channels with response times listed as N.A. No response time testing requirements apply where N.A. is listed in Table B 3.3.2-2. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the Trip Setpoint value at the sensor, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).

For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time verification is performed with the time constants set at their nominal values. The response time may be verified by a series of overlapping tests, or other verification (e.g., Ref. 8), such that the entire response time is measured.

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements (Ref. 7), provides the basis and methodology for using allocated sensor Wolf Creek - Unit 1 B 3.3.2-49 Revision 89

ENCLOSURE Page 109 of 264 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.10 (continued)

REQUIREMENTS response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

The allocations for sensor response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where response time could be affected is replacing the sensing assembly of a transmitter.

The NRC approved the use of ASME Code Case OMN-1, Alternative Rules for Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in Light-Water Reactor Plants, as an alternative to stroke time testing for motor-operated valves (Ref. 14). The parameters that must be present to achieve the analyzed response time under design basis conditions are measured to ensure the valve is capable of performing its safety function. This process verifies design basis capability, including response time, and is a significant improvement over simple stroke time measurement. This process allows the establishment of periodic valve test intervals if there is assurance that the valve will remain capable of performing its safety function throughout the interval.

ESF response times specified in Table B 3.3.2-2 which include sequential operation of RWST and VCT valves (Notes 3 and 4) are based on values assumed in the non-LOCA safety analyses. These analyses take credit for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves.

When the sequential operation of the RWST and VCT valves is not included in the response times (Note 7), the values specified are based on the LOCA analyses. The LOCA analyses take credit for injection flow regardless of the source. Verification of the response times specified in Table B 3.3.2-2 will assure that the assumptions used for the LOCA and non-LOCA analyses with respect to the operation of the VCT and RWST valves are valid.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 900 psig in the SGs.

Wolf Creek - Unit 1 B 3.3.2-50 Revision 89

ENCLOSURE Page 110 of 264 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.11 REQUIREMENTS (continued) 3.3.2.11 is the performance of a TADOT as described in SR 3.3.2.8, except that it is performed for the P-4 Reactor Trip Interlock. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint. This TADOT does not include the circuitry associated with steam dump operation since it is control grade circuitry.

SR 3.3.2.12 SR 3.3.2.12 is the performance of a monthly COT on ESFAS Function 6.h, Auxiliary Feedwater Pump Suction Transfer on Suction Pressure - Low.

A COT is performed to ensure the channel will perform the intended Function. Setpoints must be found within the Allowable Values specified in Table 3.3.2-1.

The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.2-51 Revision 89

ENCLOSURE Page 111 of 264 ESFAS Instrumentation B 3.3.2 BASES REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 7.

3. USAR, Chapter 15.

4. IEEE-279-1971.

5. 10 CFR 50.49.

6. WCNOC Nuclear Safety Analysis Setpoint Methodology for the Reactor Protection System, TR-89-0001.

7. WCAP-10271-P-A Supplement 2, Revision 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," June 1990.

8. WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996.

9. Wolf Creek Setpoint Methodology Report, SNP (KG)-492, August 29, 1984.

10. Amendment No. 43 to Facility Operating License No. NPF-42, March 29, 1991.

11. WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, October 1998.

12. 10 CFR 50.55a(b)(3)(iii), Code Case OMN-1.

13. Performance Improvement Request (PIR) 2005-2067.

Wolf Creek - Unit 1 B 3.3.2-52 Revision 89

ENCLOSURE Page 112 of 264 PAM Instrumentation B 3.3.3 BASES ACTIONS F.1 (continued)

Alternate means of monitoring Reactor Vessel Water Level and Containment Area Radiation have been developed. These alternate means may be temporarily used if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. If these alternate means are used, the Required Action is not to shut down the unit but rather to follow the directions of Specification 5.6.8, in the Administrative Controls section of the TS. Monitoring the core exit thermocouples, pressurizer level indication (BB-LI-0459A, -0460A, or -0461) and RCS subcooling monitor indication (BB-TI-1390A or B) provide an alternate means for RVLIS. These 3 parameters provide diverse information to verify there is adequate core cooling. When Containment Radiation Level (High Range) monitors are inoperable, portable survey equipment with the capability to detect gamma radiation over the range 1E-03 to 1E04 provides an alternate means (Ref. 5).

SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 REQUIREMENTS and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.

SR 3.3.3.1 Performance of the CHANNEL CHECK ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar unit instruments located throughout the unit.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

Wolf Creek - Unit 1 B 3.3.3-13 Revision 89

ENCLOSURE Page 113 of 264 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.1 (continued)

REQUIREMENTS As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.3.2 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measure parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. Containment Radiation Level (High Range) CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Whenever an RTD is replaced in Functions 2 or 3, the next required CHANNEL CALIBRATION of the RTD's is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. Whenever a core exit thermocouple is replaced in Functions 14, 15, 16, or 17, the next required CHANNEL CALIBRATION of the core exit thermocouples is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

REFERENCES 1. USAR Appendix 7A.

2. Regulatory Guide 1.97, Rev. 2, December 1980.

3. NUREG-0737, Supplement 1, TMI Action Items."

4. USAR Figure 5.1-1 (sheet 4).

5. NA 94-0089 dated May 24, 1994.

Wolf Creek - Unit 1 B 3.3.3-14 Revision 89

ENCLOSURE Page 114 of 264 Remote Shutdown System B 3.3.4 BASES ACTIONS B.1 and B.2 (continued)

If the Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.3.4.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

For the RTB Position Function, this Surveillance Requirement is met by verifying the actual position at the RTB Switchgear to the RTB indication.

For the RCP Breakers Function, this Surveillance Requirement is met by verifying the local breaker indication to the control room remote breaker indication.

As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized. Source Range Neutron Flux is de-energized in MODE 1 and in MODE 2 above the P-6 setpoint.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.4-5 Revision 89

ENCLOSURE Page 115 of 264 Remote Shutdown System B 3.3.4 BASES SURVEILLANCE SR 3.3.4.2 REQUIREMENTS (continued) SR 3.3.4.2 verifies each required Remote Shutdown System ASP control circuit and transfer switch performs the intended function. This verification is performed from the auxiliary shutdown panel. Operation of the equipment from the auxiliary shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the unit can be placed and maintained in MODE 3 from the auxiliary shutdown panel and the local stations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.4.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Notes 1 and 2 have been added to exclude the Neutron detectors (Note 1), the reactor trip breakers and RCP breakers (Note 2) from CHANNEL CALIBRATION.

Whenever an RTD is replaced in Function 5 or 6, the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 19.

2. USAR Table 7.5-2.

Wolf Creek - Unit 1 B 3.3.4-6 Revision 89

ENCLOSURE Page 116 of 264 LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS B.1 (continued)

MODES 1 - 4 and takes into account the low probability of an event requiring an LOP start occurring during this interval. When the associated DG is required to be OPERABLE in MODES 5 and 6, the Completion Time of Required Action C.1 in LCO 3.8.2, AC Sources - Shutdown, is consistent with the required times for actions requiring prompt action.

SURVEILLANCE SR 3.3.5.1 REQUIREMENTS Not Used.

SR 3.3.5.2 SR 3.3.5.2 is the performance of a TADOT. The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment. For these tests, the relay Trip Setpoints are verified and adjusted as necessary. The SR is modified by a Note that excludes verification of time delays. Testing of the time delay relays is performed as part of the CHANNEL CALIBRATION (SR 3.3.5.3). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the measured setpoint does not exceed the Allowable Value, the trip device is considered OPERABLE.

SR 3.3.5.3 SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

Calculation XX-E-009 (Ref. 3) calculates the undervoltage/degraded voltage setpoints for the NB/NG relays. The calculation also ensures adequate voltage will be present at the end use loads under minimum switchyard voltage and maximum accident loading. Calculation XX-E-009 identifies that for the degraded voltage Function, the minimum acceptable voltage for the NB01 and NB02 buses is 3756.8V. This provides additional margin below the TS Allowable Value bus voltage of 3773.8V (includes Potential Transformer error) or 107.5V at the relay (excludes Wolf Creek - Unit 1 B 3.3.5-6 Revision 89

ENCLOSURE Page 117 of 264 LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE SR 3.3.5.3 (continued)

REQUIREMENTS Potential Transformer error). For the loss of voltage Function, the minimum acceptable voltage is 3150.4V. This provides additional margin below the TS Allowable Value bus voltage of 3159.5V (includes Potential Transformer error) or 90.0V at the relay (excludes Potential Transformer error).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.5.4 SR 3.3.5.4 is the performance of the required response time verification.

This SR measures the total response time of the undervoltage relays, logic circuitry and DG start time. Response time verification acceptance criteria are:

INITIATING SIGNAL AND FUNCTION RESPONSE TIME Loss of Power

a. 4kV Bus Undervoltage -

Loss of Voltage

b. 4kV Bus Undervoltage - 78 seconds Grid Degraded Voltage The Grid Degraded Voltage Function total response time is without an SIS present.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 8.3.

2. USAR, Chapter 15.

3. Calculation XX-E-009, System NB, NG, PG Undervoltage/Degraded Voltage Relay Setpoints.

Wolf Creek - Unit 1 B 3.3.5-7 Revision 89

ENCLOSURE Page 118 of 264

ENCLOSURE Page 119 of 264 Containment Purge Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.3 REQUIREMENTS (continued) A COT is performed on each required channel to ensure the entire channel will perform the intended Function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This test verifies the capability of the instrumentation to provide the containment purge system isolation. The trip setpoint concentration value (uCi/cm3) is to be established such that the actual submersion rate would not exceed 9 mr/h in the containment building. The setpoint value may be increased up to the equivalent limits of Section 3.1 of the ODCM in accordance with the methodology and parameters in the ODCM during containment purge or vent provided the setpoint value does not exceed twice the maximum concentration activity in the containment determined by the sample analysis performed prior to each release in accordance with Table 3-1 of the ODCM.

SR 3.3.6.4 SR 3.3.6.4 is the performance of a TADOT. This test is a check of the Manual Actuation Functions. Each Manual Actuation Function is tested through the BOP ESFAS logic.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.6.5 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.6-6 Revision 89

ENCLOSURE Page 120 of 264 Containment Purge Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.6 REQUIREMENTS (continued) SR 3.3.6.6 is the performance of the required response time verification.

Response time verification acceptance criteria for the containment purge isolation instrumentation is 2 seconds. This response time acceptance criteria does not include valve closure time. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.67.

Wolf Creek - Unit 1 B 3.3.6-7 Revision 89

ENCLOSURE Page 121 of 264 CREVS Actuation Instrumentation B 3.3.7 BASES ACTIONS E.1 and E.2 (continued)

Condition E applies when the Required Action and associated Completion Time for Conditions A, B or C have not been met when irradiated fuel assemblies are being moved. Movement of irradiated fuel assemblies and CORE ALTERATIONS must be suspended immediately to reduce the risk of accidents that would require CREVS actuation. This does not preclude movement of a component to a safe position.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.7-1 REQUIREMENTS determines which SRs apply to which CREVS Actuation Functions.

SR 3.3.7.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.7.2 A COT is performed on each required channel to ensure the entire channel will perform the intended function. This test verifies the capability of the instrumentation to provide the CREVS actuation. The setpoints shall be left consistent with Note (b) of Table 3.3.7-1. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.7-6 Revision 89

ENCLOSURE Page 122 of 264 CREVS Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.3 REQUIREMENTS (continued) SR 3.3.7.3 is the performance of an ACTUATION LOGIC TEST using the BOP ESFAS automatic tester. The continuity check does not have to be performed, as explained in the Note. This SR is applied to the balance of plant actuation logic and relays that do not have circuits installed to perform the continuity check. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.7.4 SR 3.3.7.4 is the performance of a TADOT. This test is a check of the Manual Actuation Functions. Each Manual Actuation Function is tested through the BOP ESFAS.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

SR 3.3.7.5 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.7-7 Revision 89

ENCLOSURE Page 123 of 264 CREVS Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.6 REQUIREMENTS (continued) SR 3.3.7.6 is the performance of the required response time verification.

Response time verification acceptance criteria for the CREVS actuation instrumentation is 60 seconds. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.7.6 is modified by a Note stating that the radiation monitor detectors are excluded from ESF RESPONSE TIME testing. The Note is necessary because of the difficulty associated with generating an appropriate radiation monitor detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. Response time of the channel shall be verified from the detector output or input to the first electronic component in the channel.

REFERENCES 1. USAR Section 7.3.4 and Table 7.3-8.

Wolf Creek - Unit 1 B 3.3.7-8 Revision 89

ENCLOSURE Page 124 of 264

ENCLOSURE Page 125 of 264 EES Actuation Instrumentation B 3.3.8 BASES SURVEILLANCE SR 3.3.8.2 REQUIREMENTS (continued) A COT is performed on each required channel to ensure the entire channel will perform the intended function. This test verifies the capability of the instrumentation to provide the EES actuation. The setpoints shall be left consistent with the unit specific calibration procedure tolerance.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.3.8.3 SR 3.3.8.3 is the performance of an ACTUATION LOGIC TEST using the BOP ESFAS automatic tester. All possible logic combinations, with and without applicable permissive, are tested for each protection function.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The SR is modified by a Note stating that the continuity check may be excluded. This SR is applied to the balance of plant actuation logic and relays that do not have circuits installed to perform the continuity check.

SR 3.3.8.4 SR 3.3.8.4 is the performance of a TADOT. This test is a check of the manual actuation functions. Each manual actuation function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (e.g., pump starts, valve cycles, etc.).

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

SR 3.3.8.5 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measure parameter within the necessary range and accuracy. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.3.8-6 Revision 89

ENCLOSURE Page 126 of 264 EES Actuation Instrumentation B 3.3.8 BASES REFERENCES 1. 10 CFR 50.67.

2. Calculation J-G-SA02.

3. USAR Section 7.3.3 and Table 7.3-5.

4. USAR Section 15.7.4.

Wolf Creek - Unit 1 B 3.3.8-7 Revision 89

ENCLOSURE Page 127 of 264 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS B.2 (continued)

This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action. For example, this means that, during performance of Required Action B.1.1, if the flow rate is restored to within limit at 80% RTP, power does not need to be reduced below 50% RTP or 75% RTP to comply with Required Action B.2.

C.1 If the Required Actions of Conditions A or B is not met 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 at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis bounds. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.1.2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.4.1-5 Revision 89

ENCLOSURE Page 128 of 264 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE SR 3.4.1.3 REQUIREMENTS (continued) The installed flow instrumentation provides indication as a percentage of total flow rate based on the precision calorimetric heat balance. Plant procedures specify the percentage of the total flow rate required to meet the RCS total flow rate limit. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance allows the installed RCS flow instrumentation to be normalized and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate. When performing a precision heat balance, the instrumentation used for determining steam pressure, feedwater pressure, feedwater temperature, and feedwater venturi p in the calorimetric calculations shall be calibrated within 7 days prior to performing the heat balance.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 7 days after 95% RTP. This exception is appropriate since the heat balance requires the plant to be at a minimum of 95% RTP to obtain the stated RCS flow accuracies and the test is only a confirmation of SR 3.4.1.4. The Surveillance shall be performed within 7 days after reaching 95% RTP.

REFERENCES 1. USAR, Chapter 15.

2. WCAP-18083-P, Revision 0, Westinghouse Revised Thermal Design Procedure Uncertainty Calculations for the Wolf Creek Generating Station, February 2016.

Wolf Creek - Unit 1 B 3.4.1-6 Revision 89

ENCLOSURE Page 129 of 264

ENCLOSURE Page 130 of 264 RCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued)

Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed prior to entry into MODE 4. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, or inspection of the components.

ASME Code,Section XI, Appendix E (Ref. 7), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.

Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is required when RCS pressure and temperature conditions are undergoing planned changes.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

Wolf Creek - Unit 1 B 3.4.3-6 Revision 89

ENCLOSURE Page 131 of 264 RCS Loops-Modes 1 and 2 B 3.4.4 BASES APPLICABILITY The decay heat production rate is much lower than the full power heat (continued) rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

Operation in other MODES is covered by:

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

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification that each RCS loop is in operation.

Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to DNB. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Chapter 15.

Wolf Creek - Unit 1 B 3.4.4-3 Revision 89

ENCLOSURE Page 132 of 264 RCS Loops - MODE 3 B 3.4.5 BASES SURVEILLANCE SR 3.4.5.3 REQUIREMENTS (continued) Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP 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 availability to the required RCPs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.4.6.

Wolf Creek - Unit 1 B 3.4.5-6 Revision 89

ENCLOSURE Page 133 of 264

ENCLOSURE Page 134 of 264 RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.4 (continued)

REQUIREMENTS 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.4.6 Wolf Creek - Unit 1 B 3.4.6-6 Revision 89

ENCLOSURE Page 135 of 264

ENCLOSURE Page 136 of 264 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE SR 3.4.7.2 REQUIREMENTS (continued) Verifying that at least two SGs are OPERABLE by ensuring their secondary side wide range water levels are 66% ensures an alternate decay heat removal method is available via natural circulation in the event that the second RHR loop is not OPERABLE. If both RHR loops are OPERABLE, this Surveillance is not needed. The Surveillance Frequency 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 wide range water level is 66% in at least two SGs, this Surveillance is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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 Wolf Creek - Unit 1 B 3.4.7-5 Revision 89

ENCLOSURE Page 137 of 264 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE SR 3.4.7.4 (continued)

REQUIREMENTS 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 the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. 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 controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.4.6.

2. NRC Information Notice 95-35, Degraded Ability of SGs to Remove Decay Heat by Natural Circulation.

Wolf Creek - Unit 1 B 3.4.7-6 Revision 89

ENCLOSURE Page 138 of 264 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES APPLICABILITY Since LCO 3.4.8 contains Required Actions with immediate Completion (continued) Times, it is not permitted to enter LCO 3.4.8 from either LCO 3.4.7, RCS Loops - MODE 5, Loops Filled, or from MODE 6, unless the requirements of LCO 3.4.8 are met. This precludes removing the heat removal path afforded by the steam generators with the RHR System is degraded.

ACTIONS A.1 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 into the RCS, coolant with boron concentration less than required to meet the minimum 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. Boron dilution requires forced circulation from at least one RCP for proper mixing so that inadvertent criticality can be prevented.

Suspending the introduction into the RCS, 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 subcritical operations.

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 may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.4.8-3 Revision 89

ENCLOSURE Page 139 of 264 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE SR 3.4.8.2 REQUIREMENTS (continued) 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 b y verifying proper breaker alignment and power available to the RHR pump.

The Surveillance Frequency 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.

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 the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. 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 Wolf Creek - Unit 1 B 3.4.8-4 Revision 89

ENCLOSURE Page 140 of 264 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE SR 3.4.8.3 (continued)

REQUIREMENTS 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 controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.4.6.

Wolf Creek - Unit 1 B 3.4.8-5 Revision 89

ENCLOSURE Page 141 of 264 Pressurizer B 3.4.9 BASES ACTIONS C.1 and C.2 (continued)

If one group of backup pressurizer heaters are inoperable and cannot be restored in the allowed Completion Time of Required Action B.1, 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated backup pressurizer heaters are verified to be at their design rating. This may be done by energizing the heaters and measuring circuit current. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Chapter 15.

2. NUREG-0737, November 1980.

Wolf Creek - Unit 1 B 3.4.9-4 Revision 89

ENCLOSURE Page 142 of 264 Pressurizer PORVs B 3.4.11 BASES ACTIONS G.1 and G.2 (continued)

If the Required Actions of Condition F are not met, then 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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. In MODES, 3 (with any RCS cold leg temperature 368 F), 4, 5, and 6 (with the reactor vessel head on) automatic PORV OPERABILITY may be required. See LCO 3.4.12.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be opened and closed.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The Note modifies this SR by stating that it is not required to be performed with the block valve closed, in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.

SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. Operating experience has shown that these valves usually pass the Surveillance when performed at the required Inservice Testing Program frequency. The Frequency is acceptable from a reliability standpoint.

REFERENCES 1. USAR, Figure 7.2-1 (Sheet 11) and 7.6-4 (Sheets 1-3).

2. Regulatory Guide 1.32, February 1977.

3. USAR, Section 15.2.

Wolf Creek - Unit 1 B 3.4.11-7 Revision 89

ENCLOSURE Page 143 of 264

ENCLOSURE Page 144 of 264

ENCLOSURE Page 145 of 264 LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.9 REQUIREMENTS (continued) Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix G.

2. Generic Letter 88-11.

3. ASME, Boiler and Pressure Vessel Code,Section III.

4. USAR, Chapter 15.

5. 10 CFR 50, Section 50.46.

6. 10 CFR 50, Appendix K.

7. Generic Letter 90-06.

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

9. USAR, Section 5.2.2.10.

Wolf Creek - Unit 1 B 3.4.12-14 Revision 89

ENCLOSURE Page 146 of 264 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 (continued)

REQUIREMENTS gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an RCS water inventory balance.

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows). The Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is preferred when performing a proper inventory balance since calculations during non-steady state conditions must account for the changing parameters. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows. An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance. Primary to secondary LEAKAGE is determined by SR 3.4.13.2 and is not determined by an RCS water inventory balance. For determining identified LEAKAGE, identified LEAKAGE includes primary to secondary LEAKAGE as defined in Section 1.1, Definitions.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to Wolf Creek - Unit 1 B 3.4.13-5 Revision 89

ENCLOSURE Page 147 of 264 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.2 (continued)

REQUIREMENTS secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, Steam Generator Tube Integrity, should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 7. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 7).

REFERENCES 1. 10 CFR 50, Appendix A, GDC 4 and 30.

2. Regulatory Guide 1.45, May 1973.

3. USAR, Chapter 15.

4. NUREG-1061, Volume 3, November 1984.

5. 10 CFR 50.67.

6. NEI 97-06, Steam Generator Guidelines.

7. EPRI, Pressurized Water Reactor Primary-to-Secondary Leak Guidelines.

8. Standard Review Plan (SRP), Section 15.0.1.

Wolf Creek - Unit 1 B 3.4.13-6 Revision 89

ENCLOSURE Page 148 of 264 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

Testing is to be performed every 18 months, but may be extended, if the plant does not go into MODE 5 for at least 7 days. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Test pressures less than 2235 psig but greater than 150 psig are allowed for valves where higher pressures could tend to diminish leakage channel opening. Observed leakage shall be adjusted for actual pressure to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one half power.

In addition, testing must be performed once after the check valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the check valve has been reseated. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is a reasonable and practical time limit for performing this test after opening or reseating a check valve.

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Wolf Creek - Unit 1 B 3.4.14-5 Revision 89

ENCLOSURE Page 149 of 264

ENCLOSURE Page 150 of 264 RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS F.1 (continued)

With all required monitoring methods inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere particulate radioactivity monitor. The check gives reasonable confidence that the channel is operating properly.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere particulate radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner.

The test verifies the alarm setpoint and relative accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.4.15.3, SR 3.4.15.4, and SR 3.4.15.5 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

2. Regulatory Guide 1.45.

3. USAR, Section 5.2.5.

4. NUREG-609, Asymmetric Blowdown Loads on PWR Primary Systems, 1981.

Wolf Creek - Unit 1 B 3.4.15-7 Revision 89

ENCLOSURE Page 151 of 264 RCS Specific Activity B 3.4.16 BASES ACTIONS B.1 (continued)

With the DOSE EQUIVALENT XE-133 in excess of the allowed limit, DOSE EQUIVALENT XE-133 must be restored to within limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The allowed Completion Time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of an SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4c. This allowance permits entry into the applicable MODE(s), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

C.1 and C.2 If the Required Action and associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT I-131 is 60 Ci/gm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The 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 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 in Specification 1.1, Definitions, is not detected, it should be assumed to be present at the minimum detectable activity.

Wolf Creek - Unit 1 B 3.4.16-4 Revision 89

ENCLOSURE Page 152 of 264 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE SR 3.4.16.1 (continued)

REQUIREMENTS The Note modifies this SR to allow entry into and operating in MODE 4, MODE 3, and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

SR 3.4.16.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power changes when iodine spiking is more apt to occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a power change 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, is established because the iodine levels peak during this time following iodine spiking information; samples at other times would provide inaccurate results.

The Note modifies this SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

REFERENCES 1. 10 CFR 50.67.

2. Standard Review Plan (SRP), Section 15.0.1.

3. USAR Section 15.1.5.

4. USAR, Section 15.6.3.

Wolf Creek - Unit 1 B 3.4.16-5 Revision 89

ENCLOSURE Page 153 of 264 Accumulators B 3.5.1 BASES ACTIONS B.1 (continued) conditions, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-15049-A, Rev. 1 (Ref. 5).

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and RCS pressure reduced to 1000 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.

D.1 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.

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS Each accumulator valve should be verified to be fully open every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.5.1-6 Revision 89

ENCLOSURE Page 154 of 264 Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 and SR 3.5.1.3 REQUIREMENTS (continued) Borated water volume and nitrogen cover pressure are verified for each accumulator. The limit on borated water volume is equivalent to 30 %

and 70.3 % level. Only one set of non-safety channels (1 of 2) is required for water level and pressure indication. The Surveillance Frequency 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 controlled under the Surveillance Frequency Control Program. Sampling the affected accumulator within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a 70 gallon increase (approximately 8% level) 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) and the RWST has not been diluted since verifying that its boron concentration satisfies SR 3.5.4.3, because the water contained in the RWST is normally within the accumulator boron concentration requirements. This is consistent with the recommendation of NUREG-1366 (Ref. 4).

SR 3.5.1.5 Verification that power is removed from each accumulator isolation valve operator 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 controlled under the Surveillance Frequency Control Program.

This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is 1000 psig, thus allowing operational Wolf Creek - Unit 1 B 3.5.1-7 Revision 89

ENCLOSURE Page 155 of 264 ECCS - Operating B 3.5.2 BASES ACTIONS A.1 (continued) 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. This allows increased flexibility in plant operations under circumstances when components in opposite trains are inoperable.

An event accompanied by a loss of offsite power and the failure of an EDG can disable one ECCS train until power is restored. A reliability analysis (Ref. 5) 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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.

C.1 Condition A is applicable with one or more trains inoperable. The allowed Completion Time is based on the assumption that at least 100% of the ECCS flow equivalent to a single OPERABLE ECCS train is available.

With less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available, the unit is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

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 in the correct position by a power lockout isolation device 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 References 7 and 8, that can disable the function of both ECCS trains and invalidate the accident analyses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.5.2-7 Revision 89

ENCLOSURE Page 156 of 264 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.2 REQUIREMENTS (continued) 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, or securing. This SR does not apply to manual vent/drain valves, and to valves that cannot be inadvertently misaligned such as check valves. 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 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.

Wolf Creek - Unit 1 B 3.5.2-8 Revision 89

ENCLOSURE Page 157 of 264 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.3 (continued)

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

Acceptance criteria are established for the volume of accumulated gas at susceptible locations. In conjunction with or in lieu of venting, Ultrasonic Testing (UT) may be performed to verify the ECCS pumps and associated piping are sufficiently full of water. The design of the centrifugal charging pump is such that significant noncondensible gases do not collect in the pump. Therefore, it is unnecessary to require periodic pump casing venting to ensure the centrifugal charging pump will remain OPERABLE. 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 the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. If it is determined by subsequent evaluation that the ECCS 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.

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 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 controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.5.2-9 Revision 89

ENCLOSURE Page 158 of 264

ENCLOSURE Page 159 of 264 ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.7 REQUIREMENTS (continued) The position of throttle valves in the flow path is necessary for proper ECCS performance. These valves are necessary to restrict flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The ECCS throttle valves are set to ensure proper flow resistance and pressure drop in the piping to each injection point in the event of a LOCA. Once set, these throttle valves are secured with locking devices and mechanical position stops.

These devices help to ensure that the following safety analyses assumptions remain valid: (1) both the maximum and minimum total system resistance; (2) both the maximum and minimum branch injection line resistance; and (3) the maximum and minimum ranges of potential pump performance. These resistances and pump performance ranges are used to calculate the maximum and minimum ECCS flows assumed in the LOCA analyses of Reference 3.

SR 3.5.2.8 This SR requires verification that each ECCS train containment sump inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion. A visual inspection of the suction inlet piping verifies the piping is unrestricted. A visual inspection of the accessible portion of the containment sump strainer assembly verifies no evidence of structural distress or abnormal corrosion. Verification of no evidence of structural distress ensures there are no openings in excess of the maximum designed strainer opening.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

2. 10 CFR 50.46.

3. USAR, Sections 6.3 and 15.6.

4. USAR, Chapter 15, "Accident Analysis."

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

6. IE Information Notice No. 87-01.

Wolf Creek - Unit 1 B 3.5.2-11 Revision 89

ENCLOSURE Page 160 of 264 RWST B 3.5.4 BASES ACTIONS B.1 (continued) can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required. The short time limit of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains.

C.1 and C.2 If the RWST 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 at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.5.4.1 REQUIREMENTS The RWST borated water temperature should be verified to be within the limits assumed in the accident analyses band. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperatures are within the operating limits of the RWST. With ambient air temperatures within the band, the RWST temperature should not exceed the limits.

SR 3.5.4.2 The RWST water volume should be verified to be above the required minimum level ( 94% level) in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.5.4-5 Revision 89

ENCLOSURE Page 161 of 264 RWST B 3.5.4 BASES SURVEILLANCE SR 3.5.4.3 REQUIREMENTS (continued) The boron concentration of the RWST should be verified to be within the required limits. This SR verifies the boron concentration by sampling, calculation, or administrative means. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Chapter 6 and Chapter 15.

Wolf Creek - Unit 1 B 3.5.4-6 Revision 89

ENCLOSURE Page 162 of 264 Seal Injection Flow B 3.5.5 BASES ACTIONS A.1 (continued) be taken to restore the flow to below its limit. The operator has 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from the time the flow is known to be above the limit to correctly position the manual seal injection throttle valves and thus be in compliance with the accident analysis. The Completion Time minimizes the potential exposure of the plant to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow within limits.

This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.

B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification that the manual seal injection throttle valves are adjusted to give a flow within the limit ensures that proper manual seal injection throttle valve position, and hence, proper seal injection flow, is maintained. To verify acceptable seal injection flow, the following is performed; differential pressure between the charging header (PT-120) and the RCS is determined and the seal injection flow is verified to be within the limits of Figure 3.5.5-1. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.5.5-3 Revision 89

ENCLOSURE Page 163 of 264 Containment Air Locks B 3.6.2 BASES ACTIONS D.1 and D.2 (continued) 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.2.1 REQUIREMENTS Maintaining containment air locks OPERABLE requires compliance with the leakage rate test requirements of the Containment Leakage Rate Testing Program. This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The acceptance criteria were established during initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.

The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria which is applicable to SR 3.6.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.

SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.6.2-6 Revision 89

ENCLOSURE Page 164 of 264 Containment Air Locks B 3.6.2 BASES REFERENCES 1. 10 CFR 50, Appendix J, Option B.

2. USAR, Section 3.8, 6.2, and 15.

Wolf Creek - Unit 1 B 3.6.2-7 Revision 89

ENCLOSURE Page 165 of 264 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.2 (continued)

REQUIREMENTS considerations for personnel entry, or for Surveillances that require the valves to be open. The mini-purge 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 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 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 a system walkdown (which may include the use of local or remote indicators), that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. The Surveillance Frequency 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. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in 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 ALARA 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 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 Wolf Creek - Unit 1 B 3.6.3-10 Revision 89

ENCLOSURE Page 166 of 264 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.6 REQUIREMENTS (continued) Leakage integrity tests with a maximum allowable leakage rate for containment shutdown purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could develop.

This SR is modified by a Note indicting that the SR is only required to be performed when the containment shutdown purge valves blind flanges are installed.

If the blind flange is installed, leakage rate testing of the valve and its associated blind flange must be performed and following each reinstallation of the blind flange. Operating experience has demonstrated that this testing frequency is adequate to assure this penetration is leak tight.

The combined leakage rate for the containment shutdown purge supply and exhaust isolation valves, when pressurized to P a, and included with all Type B and C penetrations is less than .60 L a.

SR 3.6.3.7 For containment mini-purge and shutdown purge 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.

Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Additionally, this SR must be performed within 92 days after opening the valve. The 92 day Frequency was chosen recognizing that cycling the valve could introduce additional seal degradation (beyond that occurring to a valve that has not been opened). Thus, decreasing the interval is a prudent measure after a valve has been opened.

The SR is modified by a Note indicating that the SR is only required to be performed for the containment shutdown purge valves when the associated blind flange is removed.

The measured leakage rate for each containment mini-purge supply and exhaust isolation valve with resilient seals is less than 0.05 L a when pressurized to Pa. The combined leakage rate for the containment shutdown purge supply and exhaust isolation valves, when pressurized to Pa, and included with all Type B and C penetrations is less than .60 L a.

Wolf Creek - Unit 1 B 3.6.3-12 Revision 89

ENCLOSURE Page 167 of 264 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.8 REQUIREMENTS (continued) Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. 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 controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.

2. USAR, Figure 6.2.4-1.

3. Multi-Plant Action MPA-B020, "Containment Leakage Due to Seal Deterioration."

4. USAR, Section 6.2.4.

5. NUREG-0881, Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1, Section 6.2.3, April 1982.

6. NRC letter dated March 29, 2001, Relief Request from the Requirements of ASME Code,Section XI, Related to Code Case OMN-1 for Wolf Creek Generating Station (TAC NO. MB0982).

7. WCAP-15791-P, Rev. 1, Risk-Informed Evaluation of Extensions to Containment Isolation Valve Completion Times, April 2004.

8. License Amendment No. 190, November 3, 2010.

Wolf Creek - Unit 1 B 3.6.3-13 Revision 89

ENCLOSURE Page 168 of 264 Containment Pressure B 3.6.4 BASES ACTIONS B.1 and B.2 (continued) power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment analysis.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 6.2.

2. 10 CFR 50, Appendix K.

Wolf Creek - Unit 1 B 3.6.4-3 Revision 89

ENCLOSURE Page 169 of 264 Containment Air Temperature B 3.6.5 BASES APPLICABILITY containment average air temperature within the limit is not required in (continued) MODE 5 or 6 ACTIONS A.1 When containment average air temperature is not within the limit of the LCO, it must be restored to within limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This Required Action is necessary to return operation to within the bounds of the containment analysis. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.

B.1 and B.2 If the containment average air temperature cannot be restored to within its limit 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 3.6.5.1 REQUIREMENTS Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using data obtained from available, installed instrumentation at the following locations: a) the containment cooler inlet located near NNE wall (El.

2068'-8"), b) the containment cooler inlet located near West wall (El.

2068'-8"), c) the containment cooler inlet located near NNW wall (El.

2068'-8"), and d) the containment cooler inlet located near East wall (El.

2068'-8"). For the installed instrumentation to be considered available, the associated Containment Cooling System fan must be operating. The locations within the containment were selected to provide a representative sample of the overall containment atmosphere. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.6.5-3 Revision 89

ENCLOSURE Page 170 of 264

ENCLOSURE Page 171 of 264

ENCLOSURE Page 172 of 264 Containment Spray and Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.9 (continued)

REQUIREMENTS 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 the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. 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.

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 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 controlled under the Surveillance Frequency Control Program.

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ENCLOSURE Page 173 of 264

ENCLOSURE Page 174 of 264 Spray Additive System B 3.6.7 BASES SURVEILLANCE SR 3.6.7.2 REQUIREMENTS (continued) To provide effective iodine removal and retention, the containment spray must be an alkaline solution. Since the RWST contents are normally acidic, the volume of the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water injected. This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System. The spray additive tank site glass (ENLG0022) is utilized for meeting the SR since the control room level indicators do not provide conservative indication (Ref. 2). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.7.3 This SR provides verification of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spray solution being injected into containment is at the correct pH level. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position upon receipt of an actual or simulated actuation of a containment High-3 pressure signal.

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 controlled under the Surveillance Frequency Control Program.

SR 3.6.7.5 To ensure correct pH level is established in the borated water solution provided by the Containment Spray System, the flow rate in the Spray Additive System is verified once every 5 years. Flow of 52 gpm through the eductor test loops (supplied from the RWST) is throttled to 17 psig at the eductor inlet to simulate flow from the spray additive tank. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program Wolf Creek - Unit 1 B 3.6.7-4 Revision 89

ENCLOSURE Page 175 of 264

ENCLOSURE Page 176 of 264 MSIVs and MSIV Bypass Valves B 3.7.2 BASES SURVEILLANCE SR 3.7.2.2 REQUIREMENTS (continued) This SR verifies that each actuator train can close its respective MSIV on an actual or simulated actuation signal. The manual fast close hand switch in the control room provides an acceptable actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.2.3 This SR verifies that each MSIV bypass valve can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.2.4 This SR verifies that the closure time of each MSIV bypass valve is 15 seconds when tested pursuant to the Inservice Testing Program. This is consistent with the assumptions used in the accident and containment analyses. For the MSIV bypass valves, this Surveillance is performed routinely during plant operation (or as required for post-maintenance testing, but it may also be required to be performed upon returning the unit to operation following a refueling outage.

The Frequency for this SR is in accordance with the Inservice Testing Program.

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ENCLOSURE Page 177 of 264 MFIVs and MFRVs and MFRV Bypass Valves B 3.7.3 BASES SURVEILLANCE SR 3.7.3.2 REQUIREMENTS (continued) This SR verifies that each actuator train can close its respective MFIV on an actual or simulated actuation signal. The manual close hand switch in the control room provides an acceptable actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage in conjunction with SR 3.7.3.1. However, it is acceptable to perform this Surveillance individually. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.3.3 This SR verifies that each MFRV and MFRV bypass valve is capable of closure on an actual or simulated actuation signal. The actuation of solenoids locally at the MFRVs and MFRV bypass valves constitutes an acceptable simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage in conjunction with SR 3.7.3.1. However, it is acceptable to perform this Surveillance individually.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

REFERENCES 1. USAR, Section 10.4.7.

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

3. NUREG-1482, Revision 1, Guidelines for Inservice Testing at Nuclear Power Plants.

Wolf Creek - Unit 1 B 3.7.3-10 Revision 89

ENCLOSURE Page 178 of 264 ARVs B 3.7.4 BASES SURVEILLANCE SR 3.7.4.2 REQUIREMENTS (continued) The function of the block valve is to isolate a failed open or leaking ARV.

Cycling the block valve both closed and open demonstrates its capability to perform this function. Performance of inservice testing or use of the block valve during unit cooldown may satisfy this requirement. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 10.3.

2. USAR, Chapter 15.

3. USAR, Section 15.6.3.

4. USAR, Section 15.5.1.

Wolf Creek - Unit 1 B 3.7.4-5 Revision 89

ENCLOSURE Page 179 of 264 AFW System B 3.7.5 BASES ACTIONS D.1 (continued)

If all three AFW trains are inoperable, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with nonsafety related equipment. In such a condition, the unit should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one AFW train to OPERABLE status.

Required Action D.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to manual vent/drain valves, and to valves that cannot be inadvertently misaligned, such as check valves. 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 controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note indicating that the SR is not required to be performed for the AFW flow control valves until the AFW System is placed in standby or THERMAL POWER is above 10% RTP.

SR 3.7.5.2 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 2). Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point Wolf Creek - Unit 1 B 3.7.5-7 Revision 89

ENCLOSURE Page 180 of 264 AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.2 (continued)

REQUIREMENTS on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

Performance of inservice testing discussed in the ASME Code (Ref. 2)

(only required at 3 month intervals) satisfies this requirement. The test Frequency in accordance with the Inservice Testing Program results in testing each pump once every 3 months, as required by Reference 2.

When on recirculation, the required differential pressure for the AFW pumps (Ref. 4) when tested in accordance with the Inservice Testing Program is:

Motor Driven Pumps 1514 psid at a nominal flow of 110 gpm Turbine Driven Pump 1616.4 psid at a nominal flow of 130 gpm This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test.

SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

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 controlled under the Surveillance Frequency Control Program.

This SR includes the requirement to verify that each AFW motor-operated discharge valve limits the flow from the motor driven AFW pump to each steam generator to 320 gpm and that valves in the ESW suction flowpath actuate to the full open position upon receipt of an Auxiliary Feedwater Pump Suction Pressure-Low signal.

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ENCLOSURE Page 181 of 264 AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.4 REQUIREMENTS (continued) This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an AFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note. The Note indicates that the SR be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test.

SR 3.7.5.5 This SR verifies that the AFW is properly aligned by verifying the flow paths from the CST to each steam generator prior to entering MODE 2 after more than 30 days in MODE 5 or 6. OPERABILITY of AFW flow paths must be verified before sufficient core heat is generated that would require the operation of the AFW System during a subsequent shutdown.

The Frequency is reasonable, based on engineering judgement and other administrative controls that ensure that flow paths remain OPERABLE.

To further ensure AFW System alignment, flow path OPERABILITY is verified following extended outages to determine no misalignment of valves has occurred. This SR ensures that the flow path from the CST to the steam generators is properly aligned.

REFERENCES 1. USAR, Section 10.4.9.

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

3. NRC letter (C. Poslusny to O. Maynard) dated December 16, 1998:

Wolf Creek Generating Station - Technical Specification Bases Change, Auxiliary Feedwater System.

4. Performance Improvement Request 2002-0945.

5. Condition Report 2006-000188.

Wolf Creek - Unit 1 B 3.7.5-9 Revision 89

ENCLOSURE Page 182 of 264 CST B 3.7.6 BASES ACTIONS B.1 and B.2 (continued)

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

SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that the CST contains the required volume of cooling water. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 9.2.6.

2. USAR, Chapter 6.

3. USAR, Chapter 15.

4. Regulatory Guide 1.139, Guidance for Residual Heat Removal to Achieve and Maintain Cold Shutdown.

Wolf Creek - Unit 1 B 3.7.6-3 Revision 89

ENCLOSURE Page 183 of 264 CCW System B 3.7.7 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops - MODE 4,"

be entered if an inoperable CCW train results in an inoperable RHR lo This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.

If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.

B.1 and B.2 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.7.1 REQUIREMENTS This SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW System.

Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path provides assurance that the proper flow paths exist for CCW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to manual vent/drain valves, and to valves that cannot be inadvertently misaligned, such as check valves. 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 controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.7-3 Revision 89

ENCLOSURE Page 184 of 264 CCW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.2 REQUIREMENTS (continued) This SR verifies proper automatic operation of the CCW valves servicing safety related components or isolating the nonsafety related portion of the system on an actual or simulated actuation signal. This SR applies to the CCW valves that receive a Safety Injection signal and the RCP thermal barrier valves receiving a High CCW flow signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. 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 controlled under the Surveillance Frequency Control Program.

SR 3.7.7.3 This SR verifies proper automatic operation of the CCW pumps on an actual or simulated actuation signal. These actuation signals include Safety Injection and Loss of Power. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 9.2.2.

2. USAR, Section 6.2.

Wolf Creek - Unit 1 B 3.7.7-4 Revision 89

ENCLOSURE Page 185 of 264 ESW System B 3.7.8 BASES SURVEILLANCE SR 3.7.8.1 (continued)

REQUIREMENTS apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to being locked, sealed, or secured. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to manual vent/drain valves, and to valves that cannot be inadvertently misaligned, such as check valves.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.8.2 This SR verifies proper automatic operation of the ESW System valves servicing safety related equipment or isolating the nonsafety related portion of the system on an actual or simulated actuation signal. These actuation signals include Loss of Power, SIS, Low AFW Suction Pressure, and High Differential Test Pressure. This SR includes those valves that isolate individual components as well as those that isolate the ESW System from the SWS. The ESW System is a standby emergency system that cannot be fully actuated as part of normal testing. 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 controlled under the Surveillance Frequency Control Program.

SR 3.7.8.3 This SR verifies proper automatic operation of the ESW System pumps on an actual or simulated actuation signal. These actuation signals include SIS, Low AFW Suction Pressure, and Loss of Power. The ESW System is a standby emergency system that cannot be fully actuated as part of normal testing during normal operation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.8-4 Revision 89

ENCLOSURE Page 186 of 264 ESW System B 3.7.8 BASES REFERENCES 1. USAR, Section 9.2.1.

2. USAR, Section 15.6.

3. USAR, Section 5.4.7.

Wolf Creek - Unit 1 B 3.7.8-5 Revision 89

ENCLOSURE Page 187 of 264 UHS B 3.7.9 BASES ACTIONS A.1 and A.2 (continued)

The Completion Time of Required Action A.1 is based on engineering judgment and the fact that degradation of the main cooling lake dams structural, hydraulic, and foundation conditions is slow and significant degradation would be promptly detected and corrected prior to catastrophic failure of the main cooling lake dam.

The once per hour Completion Time of Required Action A.2 takes into consideration the increased monitoring frequency needed to ensure design basis assumptions are not exceeded in this condition.

B.1 and B.2 If the Required Actions are not completed within the associated Completion Time, or the UHS is inoperable for reasons other than Condition A, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.9.1 REQUIREMENTS This SR verifies that adequate long term (30 day) cooling can be maintained. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.9.2 This SR verifies that the ESW System is available to cool the CCW System to at least its maximum design temperature with the maximum accident or normal design heat loads for 30 days following a Design Basis Accident. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.9-3 Revision 89

ENCLOSURE Page 188 of 264 CREVS B 3.7.10 BASES ACTIONS D.1, D.2.1, and D.2.2 (continued)

An alternative to Required Action D.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

E.1 and E.2 During movement of irradiated fuel assemblies, with two CREVS trains inoperable or with one or more CREVS trains inoperable due to an inoperable CRE or CBE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the CRE. This places the unit in a condition that minimizes the accident risk. This does not preclude the movement of fuel to a safe position.

F.1 If both CREVS trains are inoperable in MODE 1, 2, 3, or 4, for reasons other than an inoperable CRE and CBE boundary (i.e., Condition B), the CREVS may not be capable of performing the intended function and the unit is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train once every month, by initiating from the control room, flow through the HEPA filters and charcoal adsorber of both the filtration and pressurization systems, provides an adequate check of this system. Monthly heater operations dry out any moisture accumulated in the charcoal from humidity in the ambient air. Each pressurization system train must be operated for 15 continuous minutes with the heaters energized. Each filtration system train need only be operated for 15 minutes continuously to demonstrate the function of the system. The 15-minute run time is based on Position C.6.1 of Reference 9. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.10-7 Revision 89

ENCLOSURE Page 189 of 264 CREVS B 3.7.10 BASES SURVEILLANCE SR 3.7.10.2 REQUIREMENTS (continued) This SR verifies that the required CREVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CREVS filter tests use the procedure guidance in Regulatory Guide 1.52, Rev. 2 (Ref. 3) in accordance with the VFTP. The VFTP includes testing the performance of the HEPA filter, charcoal absorber efficiency, minimum flow rate, and the physical properties of the activated charcoal.

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

SR 3.7.10.3 This SR verifies that each CREVS train starts and operates on an actual or simulated CRVIS. The actuation signal includes Control Room Ventilation or High Gaseous Radioactivity. The CREVS train automatically switches on an actual or simulated CRVIS into a CRVIS mode of operation with flow through the HEPA filters and charcoal adsorber banks. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE and CBE boundaries credited in the accident analysis by testing for unfiltered air inleakage past the credited envelope boundaries and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.

The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem total effective dose equivalent and the CRE occupants are protected from hazardous chemicals and smoke.

For WCGS, there is no CREVS actuation for hazardous chemical releases or smoke and there are no Surveillance Requirements that verify OPERABILITY for hazardous chemicals or smoke. This SR verifies that the unfiltered air inleakage into the CRE and CBE boundaries is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE or CBE Wolf Creek - Unit 1 B 3.7.10-8 Revision 89

ENCLOSURE Page 190 of 264 CRACS B 3.7.11 BASES ACTIONS C.1, C.2.1, and C.2.2 (continued) operation immediately. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.

An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

D.1 and D.2 In MODE 5 or 6, or during movement of irradiated fuel assemblies, with two CRACS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.

E.1 If both CRACS trains are inoperable in MODE 1, 2, 3, or 4, the CRACS may not be capable of performing its intended function. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.11.1 REQUIREMENTS Testing of the CRACS condenser heat exchangers under design conditions is impractical. This SR verifies that the heat removal capability of the CRACS air conditioning units is adequate to remove the heat load assumed in the control room during design basis accidents. This SR consists of verifying the heat removal capability of the condenser heat exchanger (either through performance testing or inspection), ensuring the proper operation of major components in the refrigeration cycle, verification of unit air flow capacity, and water flow measurement (Reference 2). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.11-3 Revision 89

ENCLOSURE Page 191 of 264 EES B 3.7.13 BASES ACTIONS D.1 and D.2 (continued)

When Required Action A.1 cannot be completed within the associated Completion Time during movement of irradiated fuel assemblies in the fuel building, the OPERABLE Emergency Exhaust System train must be started in the FBVIS mode immediately or fuel movement suspended.

This action ensures that the remaining train is OPERABLE, that no undetected failures preventing system operation will occur, and that any active failure will be readily detected.

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

E.1 If the fuel building boundary is inoperable such that a train of the Emergency Exhaust System operating in the FBVIS mode cannot establish or maintain the required negative pressure, action must be taken immediately to suspend movement of irradiated fuel assemblies in the fuel building. This does not preclude the movement of fuel assemblies to a safe position.

SURVEILLANCE SR 3.7.13.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train once every month, by initiating from the control room flow through the HEPA filters and charcoal adsorbers, provides an adequate check on this system.

Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. Systems with heaters must be operated for 15 continuous minutes with the heaters energized.

Operating heaters would not necessarily have the heating elements energized continuously for 15 minutes, but will cycle depending on the temperature. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR can be satisfied with the Emergency Exhaust System in the SIS or FBVIS lineup during testing.

The 15-minute run time is based on Position C.6.1 of Reference 10.

Wolf Creek - Unit 1 B 3.7.13-5 Revision 89

ENCLOSURE Page 192 of 264 EES B 3.7.13 BASES SURVEILLANCE SR 3.7.13.2 REQUIREMENTS (continued) This SR verifies that the required Emergency Exhaust System filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The Emergency Exhaust System filter tests are based on the guidance in References 6 and 7 in accordance with the VFTP.

The VFTP includes testing HEPA filter performance, charcoal absorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal. Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.13.3 This SR verifies that each Emergency Exhaust System train starts and operates on an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Proper completion of this SR requires testing the system in both the SIS (auxiliary building exhaust) and the FBVIS (fuel building exhaust) modes of operation.

During emergency operations the Emergency Exhaust System will automatically start in either the SIS or FBVIS lineup depending on the initiating signal. In the SIS lineup, the fans operate with dampers aligned to exhaust from the auxiliary building and prevent unfiltered leakage. In this SIS lineup, each train is capable of maintaining the auxiliary building at a negative pressure at least 0.25 inches water gauge relative to the outside atmosphere. In the FBVIS lineup, which is initiated upon detection of high radioactivity by the fuel building exhaust gaseous radioactivity monitors, the fans operate with the dampers aligned to exhaust from the fuel building to prevent unfiltered leakage. In the FBVIS lineup, each train is capable of maintaining the fuel building at a negative pressure at least 0.25 inches water gauge relative to the outside atmosphere. Normal exhaust air from the fuel building is continuously monitored by radiation detectors. One detector output will automatically align the Emergency Exhaust System in the FBVIS mode of operation.

This surveillance requirement demonstrates that each Emergency Exhaust System unit can be automatically started and properly configured to the FBVIS or SIS alignment, as applicable, upon receipt of an actual or simulated SIS signal and an FBVIS signal. It is not required that each Emergency Exhaust System unit be started from both actuation signals during the same surveillance test provided each actuation signal is tested independently.

Wolf Creek - Unit 1 B 3.7.13-6 Revision 89

ENCLOSURE Page 193 of 264 EES B 3.7.13 BASES SURVEILLANCE SR 3.7.13.4 REQUIREMENTS (continued) This SR verifies the integrity of the auxiliary building enclosure. The ability of the auxiliary building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the Emergency Exhaust System. During the SIS mode of operation, the Emergency Exhaust System is designed to maintain a slight negative pressure in the auxiliary building, to prevent unfiltered leakage. The Emergency Exhaust System is designed to maintain a negative pressure 0.25 inches water gauge with respect to atmospheric pressure at a flow rate specified in the VFTP.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.13.5 This SR verifies the integrity of the fuel building enclosure. The ability of the fuel building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the Emergency Exhaust System. During the FBVIS mode of operation, the Emergency Exhaust System is designed to maintain a slight negative pressure in the fuel building, to prevent unfiltered leakage.

The Emergency Exhaust System is designed to maintain a negative pressure 0.25 inches water gauge with respect to atmospheric pressure at a flow rate specified in the VFTP.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.13-7 Revision 89

ENCLOSURE Page 194 of 264 EES B 3.7.13 BASES REFERENCES 1. USAR, Section 6.5.1.

2. USAR, Section 9.4.2 and 9.4.3.

3. USAR, Section 15.6.5.4.

4. Regulatory Guide 1.183, Rev. 0.

5. 10 CFR 50.67.

6. ASTM D 3803-1989.

7. ANSI N510-1980.

8. Regulatory Guide 1.52, Rev. 3.

Wolf Creek - Unit 1 B 3.7.13-8 Revision 89

ENCLOSURE Page 195 of 264 Fuel Storage Pool Water Level B 3.7.15 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

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

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

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident. The water level in the fuel storage pool must be checked periodically. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling pool, and the level in the refueling pool is checked daily in accordance with SR 3.9.7.1.

Wolf Creek - Unit 1 B 3.7.15-2 Revision 89

ENCLOSURE Page 196 of 264 Fuel Storage Pool Boron Concentration B 3.7.16 BASES ACTIONS A.1, A.2.1, and A.2.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. An acceptable alternative is to verify by administrative means that the fuel storage pool verification has been performed since the last movement of fuel assemblies in the fuel storage pool. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.

If the LCO is not met while moving fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCE 1. USAR, Appendix 9.1A, The High Density Rack (HDR) Design Concept.

2. Amendment No. 120 to Facility Operating License No. NPF-42, March 22, 1999.

3. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

Wolf Creek - Unit 1 B 3.7.16-3 Revision 89

ENCLOSURE Page 197 of 264 Secondary Specific Activity B 3.7.18 BASES ACTIONS A.1 and A.2 (continued) does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.18.1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT I-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50.67.

2. USAR, Chapter 15.

3. Standard Review Plan (SRP), Section 15.0.1.

Wolf Creek - Unit 1 B 3.7.18-3 Revision 89

ENCLOSURE Page 198 of 264 SSIVs B 3.7.19 BASES ACTIONS B.1 and B.2 (continued)

If the Required Action and associated Completion Time of Condition A is not met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed at least in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.19.1 REQUIREMENTS This SR verifies the proper alignment for required automatic SSIVs in the flow path that are used to isolate the plants secondary side. The SSIV is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown (which may include the use of local or remote indicators), that valves capable of being mispositioned are in the correct position. This SR does not apply to the locked closed manual valves in the chemical injection flow path since these valves were verified to be in the correct position upon locking.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.7.19.2 This SR verifies that the isolation time of each required automatic SSIV is within limits when tested pursuant to the Inservice Testing Program. The specific limits are documented in the Inservice Testing Program. The SSIV isolation times are less than or equal to those assumed in the accident and containment analyses. The SR is performed only for required SSIVs. This Surveillance does not include verifying a closure time for the steam generator chemical injection isolation valves. An exception is made for the steam generator chemical addition injection isolation valves which are not included in the Inservice Testing Program.

These valves are passive and contain a locking device and a check valve in their flow path.

Wolf Creek - Unit 1 B 3.7.19-6 Revision 89

ENCLOSURE Page 199 of 264 SSIVs B 3.7.19 BASES SURVEILLANCE SR 3.7.19.2 (continued)

REQUIREMENTS For the required SSIVs, performance of this Surveillance is routinely done during plant operation (or as required for post-maintenance testing), but it may also be required to be performed upon returning the unit to operation following a refueling outage.

The Frequency for this SR is in accordance with the Inservice Testing Program.

SR 3.7.19.3 This SR verifies that each required automatic SSIV in the flow path is capable of closure on an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 10.4.7.

2. USAR, Section 10.4.8.

3. USAR, Section 10.3.

Wolf Creek - Unit 1 B 3.7.19-7 Revision 89

ENCLOSURE Page 200 of 264 Class 1E Electrical Equipment Air-Conditioning (A/C) System B 3.7.20 BASES ACTIONS A.1, A.2, and A.3 (continued)

The Class 1E electrical equipment A/C train must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on the low probability of an event occurring during this time period and the capability of the remaining OPERABLE Class 1E electric equipment A/C train to provide adequate area cooling for both trains of electrical equipment (with mitigating actions implemented).

If the room area temperatures are not within limits when verified once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or if the inoperable Class 1E electrical equipment A/C train cannot be restored to OPERABLE status within 30 days, Condition B must be entered.

B.1 and B.2 When the Required Actions of Condition A cannot be completed within the required Completion Times, the unit must be placed in a MODE that minimizes accident risk. To achieve this status, the unit must be placed in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

C.1 If both Class 1E electrical equipment A/C trains are inoperable in MODE 1, 2, 3 or 4, the Class 1E Electrical Equipment A/C System may not be capable of performing its intended function. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.20.1 REQUIRMENTS This SR verifies that each Class 1E electrical equipment A/C train starts and operates on an actual or simulated actuation signal. The actuation signals include the control room ventilation isolation signal (CRVIS) and actuations driven by the LOCA and shutdown sequencers. A CRVIS is generated by the inputs discussed in the LCO Bases for TS 3.3.7, CREVS Actuation Instrumentation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.7.20-4 Revision 89

ENCLOSURE Page 201 of 264 Class 1E Electrical Equipment Air-Conditioning (A/C) System B 3.7.20 BASES SURVEILLANCE SR 3.7.20.2 REQUIRMENTS (continued) Testing of the Class 1E Electrical Equipment Air Conditioning (A/C)

System condenser heat exchangers under design conditions is impractical. This SR verifies the heat removal capability of the Class 1E electrical equipment A/C trains is adequate to remove the heat load assumed in the affected rooms during design basis accidents. This SR consists of verifying the heat removal capability of the condenser heat exchanger (either through performance testing or inspection), ensuring the proper operation of major components in the refrigeration cycle, verification of unit air flow capacity, and water flow measurement. This SR is performed in the same manner as SR 3.7.11.1 (Reference 2). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 9.4.1.

2. Letter from C.F. Lyon, USNRC, to A. C. Heflin, WCNOC, Wolf Creek Generating Station - Interpretation of Technical Specification Surveillance Requirement 3.7.11.1, Verify each CRACS train has the capability to remove the assumed heat load (TAC NO. MF3665), May 28, 2014.

Wolf Creek - Unit 1 B 3.7.20-5 Revision 89

ENCLOSURE Page 202 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.1 (continued)

REQUIREMENTS independence of offsite circuits is maintained. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.

To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.

For the purposes of SR 3.8.1.2 and SR 3.8.1.7 testing, the DGs are started from standby conditions. Standby conditions for a DG mean that the diesel engine coolant and oil temperature are being maintained consistent with manufacturer recommendations.

In order to reduce stress and wear on diesel engines, the manufacturer recommends a modified start in which the starting speed of DGs is limited, warmup is limited to this lower speed, and the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 3, which is only applicable when such modified start procedures are recommended by the manufacturer.

SR 3.8.1.7 requires that the DG starts from standby conditions using one of the following signals and achieves required voltage and frequency within 12 seconds, and subsequently achieves steady state required voltage and frequency ranges:

a. Manual, or

b. Simulated loss of offsite power by itself, or

c. Safety Injection test signal.

The 12 second start requirement supports the assumptions of the design basis LOCA analysis in the USAR, Chapter 15 (Ref. 5).

Wolf Creek - Unit 1 B 3.8.1-19 Revision 89

ENCLOSURE Page 203 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (continued)

REQUIREMENTS The 12 second start requirement is not applicable to SR 3.8.1.2 (see Note 3) when a modified start procedure as described above is used. If a modified start is not used, the 12 second start requirement of SR 3.8.1.7 applies.

A minimum voltage and frequency is specified rather than an upper and lower limit because DG acceleration is likely to overshoot the upper limit initially and then go through several oscillations prior to a voltage and frequency within the stated upper and lower bounds. The time to reach steady state could exceed 12 seconds, and result in a failure of the SR.

However, on an actual emergency start, the DG would reach minimum voltage and frequency in 12 seconds at which time it would be loaded.

Application of the load will dampen the oscillations. Therefore, only specifying the minimum voltage and frequency (at which the EDG can accept load) demonstrates the necessary capability of the DG to satisfy the requirements without including a potential for failing the Surveillance.

While reaching minimum voltage and frequency (at which the DG can accept load) in 12 seconds is an immediate test of OPERABILITY, the ability of the governor and voltage regulator to achieve steady state operation, and the time to do so are important indicators of continued OPERABILITY. Therefore, the time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance. This additional monitoring and trending is part of the TR 5.5.2, Emergency Diesel Generator Reliability Program and is not considered part of the SR. (Reference 14)

Since SR 3.8.1.7 requires a 12 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This is the intent of Note 1 of SR 3.8.1.2.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.1.3 This Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to the equivalent of the maximum expected accident loads and aligned to Wolf Creek - Unit 1 B 3.8.1-20 Revision 89

ENCLOSURE Page 204 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 (continued)

REQUIREMENTS provide standby power to the associated emergency buses. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source. The DG shall be operated continuously for the 60 minute time period per the guidance of Regulatory Guide 1.9, Position 2.2.2 (Ref. 3).

Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.

The 0.8 value is the design rating of the machine, while the 1.0 is an operational limitation to ensure circulating currents are minimized. The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.

The DG is considered OPERABLE during performance of the Surveillance, i.e., while it is paralleled to the offsite power source, consistent with the Technical Evaluation (i.e., Section 4.0) contained in the Safety Evaluation provided for Amendment No. 154 (Reference 17).

This includes consideration of the potential challenges to the DG, its response to a LOCA and/or a loss of offsite power, and appropriate operator actions to restore the DG.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Momentary power factor transients outside the normal range are acceptable during this surveillance since no power factor requirements are established by this SR. Note 3 indicates that this Surveillance should be conducted on only one DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.

SR 3.8.1.4 This SR provides verification that, with the DG in a standby condition, the fuel oil transfer pump starts on low level in the day tank standpipe and shuts down on high level in the day tank standpipe to automatically maintain the day tank fuel oil level above the DG fuel headers. The fuel Wolf Creek - Unit 1 B 3.8.1-21 Revision 89

ENCLOSURE Page 205 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.4 (continued)

REQUIREMENTS oil standpipe must have adequate level to keep the fuel oil supply header to the engine injector pumps full, so that the engine can meet the required 12 second start time. The minimum fuel oil free surface elevation is required to be at least 86 inches from the bottom (outside diameter) of the tank. The transfer pump start/stop setpoints are controlled to maintain level in the standpipe in order to ensure there is sufficient fuel to meet the 12 second start requirement for the DG. This level also ensures adequate fuel oil for a minimum of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of DG operation at full load plus 10%.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day tanks eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR is for preventative maintenance. The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during the performance of this Surveillance.

SR 3.8.1.6 This Surveillance demonstrates that each required fuel oil transfer pump operates and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.8.1-22 Revision 89

ENCLOSURE Page 206 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.7 REQUIREMENTS See SR 3.8.1.2.

SR 3.8.1.8 Not Used.

SR 3.8.1.9 Not Used.

SR 3.8.1.10 This Surveillance demonstrates the DG capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits.

The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions. This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection. While the DG is not expected to experience this transient during an event and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.

The DG is considered OPERABLE while it is paralleled to the offsite power source, consistent with the Technical Evaluation (i.e., Section 4.0) contained in the Safety Evaluation provided for Amendment No. 154 (Reference 17). This includes consideration of the potential challenges to the DG, its response to a LOCA and/or a loss of offsite power, and appropriate operator actions to restore the DG.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR has been modified by a Note. The Note ensures that the DG is tested under load conditions that are as close to design basis conditions as possible. When synchronized with offsite power, testing should be of the actual inductive loading a DG would see under design basis accident conditions. Under certain conditions, however, the Note allows Wolf Creek - Unit 1 B 3.8.1-23 Revision 89

ENCLOSURE Page 207 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.11 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

The Note 2 restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post-work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

SR 3.8.1.12 This Surveillance demonstrates that the DG automatically starts and achieves the required voltage and frequency within the specified time (12 seconds) from the design basis actuation signal (LOCA signal) and operates for 5 minutes. The 5 minute period provides sufficient time to demonstrate stability. SR 3.8.1.12.d and SR 3.8.1.12.e ensure that permanently connected loads and emergency loads are energized from the offsite electrical power system on an ESF signal without loss of offsite power.

The requirement to verify the connection of permanent and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of Wolf Creek - Unit 1 B 3.8.1-25 Revision 89

ENCLOSURE Page 208 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)

REQUIREMENTS these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, ECCS injection valves are not desired to be stroked open, or high pressure injection systems are not capable of being operated at full flow, or RHR systems performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.

This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.

The Note 2 restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post-work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Wolf Creek - Unit 1 B 3.8.1-26 Revision 89

ENCLOSURE Page 209 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.13 REQUIREMENTS This Surveillance demonstrates that DG noncritical protective functions are bypassed on a loss of voltage signal concurrent with an ESF actuation test signal. The noncritical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.1.14 Regulatory Guide 1.9, Rev. 3, (Ref. 3), requires demonstration once per 18 months that the DGs can start and run continuously at full load capability for an interval of not less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of which is at a load not greater than 110% of the continuous duty rating (short-time rated load) and the remainder of the time at a load equivalent to the continuous duty rating (continuous rated load) of the DG. The short-time rated load and the continuous rated load may be applied in either order.

The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.

The load band is provided to avoid routine overloading of the DG.

Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.

Administrative controls for performing this SR in MODES 1 or 2, with the DG connected to an offsite circuit, ensure or require that:

a. Weather conditions are conducive for performing this SR.

b. The offsite power supply and switchyard conditions are conducive for performing this SR, which includes ensuring that switchyard access is restricted and no elective maintenance within the switchyard is performed.

c. No equipment or systems assumed to be available for supporting the performance of the SR are removed from service.

Wolf Creek - Unit 1 B 3.8.1-27 Revision 89

ENCLOSURE Page 210 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.14 (continued)

REQUIREMENTS The DG is considered OPERABLE during performance of the Surveillance, i.e., while it is paralleled to the offsite power source, consistent with the Technical Evaluation (i.e., Section 4.0) contained in the Safety Evaluation provided for Amendment No. 154 (Reference 17).

This includes consideration of the potential challenges to the DG, its response to a LOCA and/or a loss of offsite power, and appropriate operator actions to restore the DG.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This Surveillance is modified by two Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients outside the power factor range will not invalidate the test. Note 2 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible. When synchronized with offsite power, testing should be performed at a power factor o the actual inductive loading a DG would see under design basis accident conditions. Under certain conditions, however, Note 2 allows the Surveillance to be conducted se conditions occur when grid voltage is high, an 0.9 results in voltages on the emergency busses that are too high. Under these conditions, the power factor should be maintained as close as practicable to 0.9 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.9 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained as close as practicable to 0.9 without exceeding the DG excitation limits.

SR 3.8.1.15 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 12 seconds. The 12 second time is derived from the requirements of the accident analysis to respond to a design basis large break LOCA. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.8.1-28 Revision 89

ENCLOSURE Page 211 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.15 (continued)

REQUIREMENTS (continued) This SR is modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. The requirement that the diesel has operated for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at full load conditions prior to performance of this Surveillance is based on manufacturer recommendations for achieving hot conditions. Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.

SR 3.8.1.16 As required by Regulatory Guide 1.9, Rev. 3 (Ref. 3), this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and the DG can be returned to ready to load status when offsite power is restored. It also ensures that the autostart logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in ready to load status when the DG is at rated speed and voltage, the output breaker is open and can receive a close signal on bus undervoltage, and the load sequence timers are reset.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

The restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post-work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a Wolf Creek - Unit 1 B 3.8.1-29 Revision 89

ENCLOSURE Page 212 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.16 (continued)

REQUIREMENTS plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in MODE 1, 2, 3 or 4. Risk insights or deterministic methods may be used for this assessment.

SR 3.8.1.17 Demonstration of the test mode (parallel mode) override ensures that the DG availability under accident conditions will not be compromised as the result of testing and the DG will automatically reset to ready to load operation if a Safety Injection actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open. These provisions for automatic switchover are required by IEEE-308 (Ref. 13),

paragraph 6.2.6(2).

The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.12. The intent in the requirement associated with SR 3.8.1.17.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.

This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

The restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post-work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated Wolf Creek - Unit 1 B 3.8.1-30 Revision 89

ENCLOSURE Page 213 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.17 (continued)

REQUIREMENTS independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

SR 3.8.1.18 Under accident and loss of offsite power conditions loads are sequentially connected to the bus by the LSELS. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. The 10% load sequence time interval tolerance ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated. Reference 2 provides a summary of the automatic loading of ESF buses.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

The restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post-work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Wolf Creek - Unit 1 B 3.8.1-31 Revision 89

ENCLOSURE Page 214 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.19 REQUIREMENTS In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.

This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.11, during a loss of offsite power actuation test signal in conjunction with an ESF actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.

The Note 2 restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post-work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Wolf Creek - Unit 1 B 3.8.1-32 Revision 89

ENCLOSURE Page 215 of 264 AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.20 REQUIREMENTS (continued) This Surveillance demonstrates that the DG starting independence has not been compromised. Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil temperature maintained consistent with manufacturer recommendations.

SR 3.8.1.21 SR 3.8.1.21 is the performance of an ACTUATION LOGIC TEST using the LSELS automatic tester for each load shedder and emergency load sequencer train except that the continuity check does not have to be performed, as explained in the Note. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. USAR, Chapter 8.

3. Regulatory Guide 1.9, Rev. 3.

4. USAR, Chapter 6.

5. USAR, Chapter 15.

6. Regulatory Guide 1.93, Rev. 0, December 1974.

7. Generic Letter 84-15, "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," July 2, 1984.

8. 10 CFR 50, Appendix A, GDC 18.

9. Regulatory Guide 1.108, Rev. 1, August 1977.

10. Regulatory Guide 1.137, Rev. 0, January 1978.

Wolf Creek - Unit 1 B 3.8.1-33 Revision 89

ENCLOSURE Page 216 of 264 Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 (continued)

REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.3.2 This Surveillance ensures that sufficient lube oil inventory is available to support at least 7 days of full load operation for each DG. The 750 gal requirement is based on the DG manufacturer consumption values for the run time of the DG. The required inventory is contained entirely within the engine crankcase. There are several methods available to verify the lube oil volume. The preferred method is to verify normal levels on the dipstick. Other, indirect, methods such as the local level indicator or the absence of a low level alarm are acceptable as secondary methods.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.3.3 The tests listed below (in accordance with the Diesel Fuel Oil Testing Program in Specification 5.5.13) are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks.

These tests are to be conducted prior to adding the new fuel to the storage tank(s), but in no case is the time between receipt of new fuel and conducting the tests to exceed 31 days. The tests, limits, and applicable ASTM Standards are as follows:

a. Sample the new fuel oil in accordance with ASTM D4057-81 (Ref. 6);

Wolf Creek - Unit 1 B 3.8.3-6 Revision 89

ENCLOSURE Page 217 of 264 Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES SURVEILLANCE SR 3.8.3.3 (continued)

REQUIREMENTS The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.

SR 3.8.3.4 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available. The system design requirements provide for a minimum of five engine start cycles without recharging. A start cycle is defined as 3 seconds of cranking time or approximately 2 to 3 engine revolutions. The pressures specified in this SR are intended to reflect the lowest value at which the five starts can be accomplished with air supplied from one or two receivers.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.3.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks once every 31 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR is for preventive maintenance. The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.

Wolf Creek - Unit 1 B 3.8.3-8 Revision 89

ENCLOSURE Page 218 of 264 DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.1 (continued)

REQUIREMENTS charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or a battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. This SR applies only to those chargers connected to a battery bank and bus. (Ref. 12)

SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each intercell, and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance. The visual inspection is to detect corrosion in cell post connection area; corrosion outside the connection area is not an OPERABILITY concern and would not require measuring resistance.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the OPERABILITY of the battery (its ability to perform its design function.)

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of connections provide an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The anticorrosion material is used Wolf Creek - Unit 1 B 3.8.4-5 Revision 89

ENCLOSURE Page 219 of 264 DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.4 and SR 3.8.4.5 (continued)

REQUIREMENTS to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection. The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR provided visible corrosion is removed during performance of SR 3.8.4.4.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.4.6 This SR requires that each battery charger be capable of supplying 300 amps and 128.4 V for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. These requirements are based on the design rating of the chargers (Ref. 4). According to Regulatory Guide 1.32 (Ref. 10), the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.4.7 A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length should correspond to the design duty cycle requirements as specified in Reference 4.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.8.4-6 Revision 89

ENCLOSURE Page 220 of 264 DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.8 (continued)

REQUIREMENTS The manufacturer recommends that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. However, based on discussions with the NRC associated with the AT&T round cell batteries (Reference 13), the SR specifies a battery capacity of 85%. If battery capacity is below 85% of the manufacturers rating, the battery is to be replaced.

The battery capacity is determined using the manufacturers minimum lifetime rating. Adverse trends in the battery capacity identified during the performance of this SR are evaluated in accordance with the corrective action program.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 18 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 9), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is 10% below the manufacturer's rating.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. Regulatory Guide 1.6, March 10, 1971.

3. IEEE-308-1978.

4. USAR, Chapter 8.

5. IEEE-485-1983, June 1983.

6. USAR, Chapter 6.

7. USAR, Chapter 15.

8. Regulatory Guide 1.93, December 1974.

Wolf Creek - Unit 1 B 3.8.4-8 Revision 89

ENCLOSURE Page 221 of 264 DC Sources - Operating B 3.8.4 BASES REFERENCES 9. IEEE-450-1995.

(continued)

10. Regulatory Guide 1.32, February 1977.

11. .NRC letter (J. Stone to O. Maynard) dated February 10, 1997:

Wolf Creek Generating Station - Amendment No. 104 to Facility Operating License No. NPF-42.

12. NRC Inspection Report 50-482/98-12, Paragraph e.16.

Wolf Creek - Unit 1 B 3.8.4-9 Revision 89

ENCLOSURE Page 222 of 264 Battery Cell Parameters B 3.8.6 BASES ACTIONS B.1 (continued)

With one or more batteries with one or more battery cell parameters outside the Category C limit for any connected cell, sufficient capacity to supply the maximum expected load requirement is not assured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below 60 F, are also cause for immediately declaring the associated DC electrical power subsystem inoperable. IEEE 450 suggests that representative cells be interpreted to mean every sixth cell.

SURVEILLANCE SR 3.8.6.1 REQUIREMENTS This SR verifies that Category A battery cell parameters are consistent with IEEE-450 (Ref. 3), which recommends regular battery inspections including voltage, electrolyte level, temperature, level corrected specific gravity, and electrolyte temperature of pilot cells.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.6.2 Within 7 days of a battery discharge < 110 V or a battery overcharge

> 150 V, the battery must be demonstrated to meet Category B limits.

Transients, such as motor starting transients, which may momentarily cause battery voltage to drop to < 110 V, do not constitute a significant battery discharge provided the battery terminal voltage and float current return to pre-transient values. This inspection is also consistent with IEEE-450 (Ref. 3), which recommends special inspections following a severe discharge or overcharge, to ensure that no significant degradation of the battery occurs as a consequence of such discharge or overcharge.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.8.6.3 This Surveillance verification that the average temperature of representative cells is 60 F, is consistent with a recommendation of IEEE-450 (Ref. 3), that states that the temperature of electrolytes in representative cells should be determined on a quarterly basis. IEEE 450 suggests that representative cells be interpreted to mean every sixth cell.

Wolf Creek - Unit 1 B 3.8.6-3 Revision 89

ENCLOSURE Page 223 of 264 Battery Cell Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.3 (continued)

REQUIREMENTS Lower than normal temperatures act to inhibit or reduce battery capacity.

This SR ensures that the operating temperatures remain within an acceptable operating range. This limit is based on manufacturer recommendations.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Table 3.8.6-1 This table delineates the limits on battery cell parameters for three different categories. The meaning of each category is discussed below.

Category A defines the normal parameter limit for each designated pilot cell in each battery. The cells selected as pilot cells are those whose temperature, voltage, and electrolyte specific gravity approximate the condition of the entire battery.

The Category A limits specified for electrolyte level are based on manufacturer recommendations and are consistent with the guidance in IEEE-450 (Ref. 3), with the extra 1/4 inch allowance above the high level indication for operating margin to account for temperatures and charge effects. In addition to this allowance, footnote a to Table 3.8.6-1 permits the electrolyte level to be above the specified maximum level during equalizing charge, provided it is not overflowing. According to the manufacturer, the electrolyte level may remain high for up to two weeks following charging. These limits ensure that the plates suffer no physical damage, and that adequate electron transfer capability is maintained in the event of transient conditions. IEEE-450 (Ref. 3) recommends that electrolyte level readings should be made only after the battery has been at float charge for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The Category A limit specified for float voltage is 2.14 V per cell. This value is based on the recommendations of IEEE-450 (Ref. 3), which states that the float voltage value should be the manufacturers recommended minimum cell voltage that will prevent reducing the life expectancy of the cells. The value of 2.14 V was obtained from the manufacturer for Category A and B.

The Category A limit specified for specific gravity for each pilot cell is 1.200 (0.015 below the manufacturer fully charged nominal specific gravity or a battery charging current that had stabilized at a low value).

Wolf Creek - Unit 1 B 3.8.6-4 Revision 89

ENCLOSURE Page 224 of 264 Inverters - Operating B 3.8.7 BASES ACTIONS A.1 With a required inverter inoperable, its associated AC vital bus is inoperable until it is re-energized from its bypass constant voltage transformer or the bypass constant voltage transformer of the respective spare inverter. The bypass constant voltage transformers are powered from a Class 1E bus.

For this reason a Note has been included in Condition A requiring the entry into the Conditions and Required Actions of LCO 3.8.9, "Distribution Systems - Operating," with any vital bus de-energized. This ensures that the vital bus is re-energized within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Required Action A.1 allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to fix the inoperable inverter or place the associated train spare inverter in service. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> limit is based upon engineering judgment, taking into consideration the time required to repair an inverter and the additional risk to which the unit is exposed because of the inverter inoperability. This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the AC vital bus is powered from its bypass constant voltage transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the AC vital buses is the preferred source for powering instrumentation trip setpoint devices.

B.1 and B.2 If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter. The verification of proper voltage output ensures that the required power is readily available for the instrumentation of the RPS and ESFAS connected to the AC vital buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

Wolf Creek - Unit 1 B 3.8.7-3 Revision 89

ENCLOSURE Page 225 of 264 Distribution Systems - Operating B 3.8.9 BASES ACTIONS F.1 (continued)

With two trains with inoperable distribution subsystems that result in a loss of safety function, adequate core cooling, containment OPERABILITY and other vital functions for DBA mitigation would be compromised, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.8.9.1 REQUIREMENTS This Surveillance verifies that the AC, DC, and AC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus.

The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 15.

3. Regulatory Guide 1.93, December 1974.

Wolf Creek - Unit 1 B 3.8.9-8 Revision 89

ENCLOSURE Page 226 of 264 Distribution Systems - Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the required AC, DC, and AC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Chapter 6.

2. USAR, Chapter 15.

Wolf Creek - Unit 1 B 3.8.10-6 Revision 89

ENCLOSURE Page 227 of 264 Boron Concentration B 3.9.1 BASES ACTIONS A.3 (continued)

Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentration in the filled portions of the RCS and the refueling canal, that have direct access to the reactor vessel, is within the COLR limits. The boron concentration of the coolant in each required volume is determined periodically by chemical analysis.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

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

Wolf Creek - Unit 1 B 3.9.1-4 Revision 89

ENCLOSURE Page 228 of 264 Unborated Water Source Isolation Valves B 3.9.2 BASES SURVEILLANCE SR 3.9.2.1 (continued)

REQUIREMENTS SR 3.9.1.1. This Surveillance demonstrates that the valves are closed through a system walkdown (which may include the use of local or remote indicators). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 15.4.6.

2. NUREG-0800, Section 15.4.6.

Wolf Creek - Unit 1 B 3.9.2-3 Revision 89

ENCLOSURE Page 229 of 264 Nuclear Instrumentation B 3.9.3 BASES ACTIONS B.1 (continued)

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

B.2 With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and boron concentration changes inconsistent with Required Action A.2 are not to be made, the core reactivity condition is stabilized until the source range 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.

The Completion Time of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS SR 3.9.3.1 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 controlled under the Surveillance Frequency Control Program.

SR 3.9.3.2 SR 3.9.3.2 is the performance of a CHANNEL CALIBRATION. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The source range neutron detectors are maintained based on manufacturers Wolf Creek - Unit 1 B 3.9.3-3 Revision 89

ENCLOSURE Page 230 of 264 Nuclear Instrumentation B 3.9.3 BASES TECHNICAL SR 3.9.3.2 (continued)

SURVEILLANCE REQUIREMENTS recommendations. The Surveillance Frequency 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. NRC letter (J. Stone to O. Maynard) dated October 3, 1997:

Wolf Creek Generating Station - Technical Specification Bases Change, Source Range Nuclear Instruments Power Supply Requirements.

3. Engineering Disposition for WO 11-339015-002, Changes to TRM 3.3.15, March 21, 2011.

4. PIR 2004-1625, Gamma-Metrics Detectors for Core Alterations, October 5, 2005.

Wolf Creek - Unit 1 B 3.9.3-4 Revision 89

ENCLOSURE Page 231 of 264 Containment Penetrations B 3.9.4 BASES ACTIONS A.1 and A.2 (continued) including the containment purge isolation valve not capable of automatic actuation, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. For the open purge isolation valves, this Surveillance will ensure that each valve is not blocked from closing and each valve operator has motive power by demonstrating that each valve actuates to its isolation position. Containment penetrations that are open under administrative controls are not required to meet the SR during the time the penetrations are open.

The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.9.4.2 This Surveillance demonstrates that the necessary hardware, tools, and equipment are available to install the equipment hatch. The equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in the opening. The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions.

Wolf Creek - Unit 1 B 3.9.4-5 Revision 89

ENCLOSURE Page 232 of 264 Containment Penetrations B 3.9.4 BASES SURVEILLANCE SR 3.9.4.2 (continued)

REQUIREMENTS The Surveillance is modified by a Note which only requires that the Surveillance be met for an open equipment hatch. If the equipment hatch is installed in its opening, the availability of the means to install the hatch is not required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.9.4.3 This Surveillance demonstrates that each containment purge isolation valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surveillances will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

REFERENCES 1. Amendment No. 74 to Wolf Creek Generating Station Operating License NPF-42, dated July 7, 1994.

2. USAR, Section 15.7.4.

3. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

4. Amendment No. 95 to Wolf Creek Generating Station Operating License NPF-42, dated February 28, 1996.

5. Configuration Change Package 7784.

6. Amendment No. 135 to Wolf Creek Generating Station Operating License NPF-42, dated September 12, 2000.

Wolf Creek - Unit 1 B 3.9.4-6 Revision 89

ENCLOSURE Page 233 of 264 RHR and Coolant Circulation - High Water Level B 3.9.5 BASES ACTIONS A.4 (continued)

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

SURVEILLANCE SR 3.9.5.1 REQUIREMENTS 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 Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.9.5.2 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 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 the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. If it is determined by Wolf Creek - Unit 1 B 3.9.5-4 Revision 89

ENCLOSURE Page 234 of 264 RHR and Coolant Circulation - High Water Level B 3.9.5 BASES SURVEILLANCE SR 3.9.5.2 (continued)

REQUIREMENTS 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 controlled under the Surveillance Frequency Control Program.

REFERENCES 1. USAR, Section 5.4.7.

2. SAP-06-113, Loss of RHR Analysis with the Refuel Cavity Flooded and Upper Internals Installed, November 16, 2006.

Wolf Creek - Unit 1 B 3.9.5-5 Revision 89

ENCLOSURE Page 235 of 264 RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES SURVEILLANCE SR 3.9.6.1 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 and to prevent thermal and boron stratification in the core. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.9.6.2 Verification that the required pump is OPERABLE ensures that an additional 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 controlled under the Surveillance Frequency Control Program.

SR 3.9.6.3 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.

Wolf Creek - Unit 1 B 3.9.6-4 Revision 89

ENCLOSURE Page 236 of 264 RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES SURVEILLANCE SR 3.9.6.3 (continued)

REQUIREMENTS 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 the accumulated gas is eliminated or brought within the acceptance criteria limits during performance of the Surveillance, the Surveillance is met and past system OPERABILITY is evaluated under the Corrective Action Program. 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 controlled under the Surveillance Frequency Control Program.

1. USAR, Section 5.4.7.

2. Generic Letter No. 88-17, Loss of Decay Heat Removal.

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ENCLOSURE Page 253 of 264 RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO b. Core outlet temperature is maintained at least 10 F below (continued) 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 50 F above each of the RCS cold leg temperatures before the start of a reactor coolant pump (RCP) with any RCS cold leg temperature 368 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.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required. When both RHR loops (or trains) are required to be OPERABLE, the associated Component Cooling Water (CCW) train is required to be capable of performing its related support function(s). The heat sink for the CCW System is normally provided by the Service Water System or Essential Service Water (ESW) System, as determined by system availability. In MODES 5 and 6, one Diesel Generator (DG) is required to be OPERABLE per LCO 3.8.2, AC Sources - Shutdown. The same ESW train is required to be capable of performing its related support function(s) to support DG OPERABILITY. A SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE. 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. The RCS loops are considered filled when the RCS is fully vented. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side wide range water level of at least two SGs is required to be 66%.

Operation in other MODES is covered by:

Wolf Creek - Unit 1 B 3.4.7-3 Revision 90

ENCLOSURE Page 254 of 264 RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES LCO Note 1 permits all RHR pumps to be removed from operation for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

(continued) 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 the SDM of LCO 3.1.1 is maintained or draining operations when RHR forced flow is stopped. The Note requires reactor vessel water level be above the vessel flange to ensure the operating RHR pump will not be intentionally deenergized during mid-loop operations.

Note 2 allows one RHR loop to be inoperable for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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. The heat sink for the CCW System is normally provided by the Service Water System or Essential Service Water (ESW) System, as determined by system availability. In MODES 5 and 6, one Diesel Generator (DG) is required to be OPERABLE per LCO 3.8.2, AC Sources - Shutdown. The same ESW train is required to be capable of performing its related support function(s) to support DG OPERABILITY. Management of gas voids is important to RHR OPERABILITY.

APPLICABILITY In MODE 5 with loops not filled, this LCO requires core heat removal and coolant circulation by the RHR System. The RCS loops are considered not filled when the RCS is no longer capable of being pressurized. One RHR loop provides sufficient capability for this purpose. However, one additional RHR loop is 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 3.4.6, "RCS Loops - MODE 4";

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

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level" (MODE 6).

Wolf Creek - Unit 1 B 3.4.8-2 Revision 90

ENCLOSURE Page 255 of 264 Containment Isolation Valves B 3.6.3 BASES APPLICABLE The LOCA and rod ejection offsite dose analyses assumes leakage from SAFETY ANALYSES the containment at a maximum leak rate of 0.20 percent of the (continued) containment volume per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 0.10 percent of the containment volume per day for the duration of the accident.

The single failure criterion required to be imposed in the conduct of plant safety analyses was considered in the original design of the 18 inch containment mini-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. The inboard and outboard isolation valves on each line are provided with independent electrical power sources to solenoids that open the pneumatically operated spring closed actuators. The actuators fail closed on the loss of power or air.

This arrangement was designed to preclude common mode failures from disabling both valves on a purge line.

The 36 inch purge valves may be unable to close against the buildup of pressure following a LOCA. Therefore, each of the purge valves is required to remain sealed closed or closed and blind flange installed during MODES 1, 2, 3, and 4. The Containment Shutdown Purge 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(c)(2)(ii).

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 DBA.

The automatic power operated isolation valves are required to have isolation times within limits and to actuate on an automatic isolation signal.

The 36 inch containment purge supply and exhaust valves must be maintained sealed closed or closed and blind flange installed. The valves covered by this LCO are listed along with their associated stroke times in the USAR (Ref. 2).

The normally closed containment 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 2.

Wolf Creek - Unit 1 B 3.6.3-3 Revision 90

ENCLOSURE Page 256 of 264 Containment Isolation Valves B 3.6.3 BASES ACTIONS C.1 and C.2 (continued)

In the event one containment isolation valve in two or more separate penetration flow paths is inoperable, except for purge valve leakage not within limit, all but one of the affected penetration flow path(s) 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 with flow through the valve secured. For a penetration flow path isolated in accordance with C.1, the device used to isolate the penetration should be the closest available one to containment.

Required Action C.1 must be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. For the penetration flow paths isolated in accordance with Required Action C.1, the affected penetration(s) must be verified to be isolated on a periodic basis per Required Action A.2, which remains in effect. This periodic verification is necessary to assure that the penetrations requiring isolation following an accident are isolated. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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.

This Condition is applicable when multiple containment isolation valves in separate flow paths are inoperable. For subsequent containment isolation valve inoperabilities, the Required Action and Completion Time continue to apply to each additional containment isolation valve inoperability, with the Completion Time based on each subsequent entry into the Condition consistent with Note 2 to the ACTIONS Table (e.g., for each entry into the Condition). The containment isolation valve(s) inoperable as a result of that entry shall meet the Required Action and Completion Time.

D.1, D.2, and D.3 In the event one or more containment shutdown or mini-purge 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 or closed manual valve (this includes power operated valves with power removed) or blind flange. A containment shutdown purge or mini-purge valve with resilient seals utilized to satisfy Required Action D.1 must have been demonstrated to meet the leakage requirements of SR 3.6.3.6 or SR 3.6.3.7. The specified Completion Time is reasonable, considering that one containment purge valve remains closed so that a gross breach of containment does not exist.

Wolf Creek - Unit 1 B 3.6.3-7 Revision 90

ENCLOSURE Page 257 of 264 Containment Isolation Valves B 3.6.3 BASES ACTIONS E.1 and E.2 (continued)

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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 36 inch containment shutdown purge supply and exhaust valve is required to be verified sealed closed or closed and blind flange installed.

Each 36 inch containment shutdown purge supply and exhaust valve inside containment must be verified sealed closed or blind flange installed prior to entering MODE 4 from MODE 5, if the surveillance has not been performed in the previous 92 days. This Surveillance is designed to ensure that a gross breach of containment is not caused by an inadvertent or spurious opening of a containment shutdown purge valve.

Detailed analysis of these valves failed to conclusively demonstrate their ability to close during a LOCA in time to limit offsite doses. Therefore, these valves are required to be in the sealed closed position or closed and blind flange installed during MODES 1, 2, 3, and 4. A containment shutdown purge 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. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.6.3.2 This SR ensures that the mini-purge valves are closed as required or, if open, open for an allowable reason. If a mini-purge 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 mini-purge valves are open for the reasons stated.

The valves may be opened for pressure control, ALARA or air quality Wolf Creek - Unit 1 B 3.6.3-9 Revision 90

ENCLOSURE Page 258 of 264 AFW System B 3.7.5 BASES ACTIONS C.1 (continued) two steam supply lines to the turbine driven AFW pump. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable, based on redundant capabilities afforded by the AFW System, time needed for repairs, and the low probability of a DBA occurring during this time period.

The second Completion Time for Required Action C.1 establishes a limit on the maximum time allowed for any combination of Conditions to be inoperable during any continuous failure to meet this LCO.

The 10 day Completion Time provides a limitation time allowed in this specified Condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and C, or B and C are entered concurrently. The AND connector between 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 10 days dictates that both Completion Times apply simultaneously, and the more restrictive must be met.

D.1 and D.2 When Required Action A.1, B.1, or C.1 cannot be completed within the required Completion Time, or if two AFW trains are inoperable, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

E.1 If all three AFW trains are inoperable, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown, and only limited means for conducting a cooldown with nonsafety related equipment. In such a condition, the unit should not be perturbed by any action, including a power change, that might result in a trip. The seriousness of this condition requires that action be started immediately to restore one AFW train to OPERABLE status.

Required Action E.1 is modified by a Note indicating that all required MODE changes or power reductions are suspended until one AFW train is restored to OPERABLE status. In this case, LCO 3.0.3 is not applicable because it could force the unit into a less safe condition.

Wolf Creek - Unit 1 B 3.7.5-7 Revision 90

ENCLOSURE Page 259 of 264 AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to manual vent/drain valves, and to valves that cannot be inadvertently misaligned, such as check valves. 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 controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note indicating that the SR is not required to be performed for the AFW flow control valves until the AFW System is placed in standby or THERMAL POWER is above 10% RTP.

SR 3.7.5.2 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (Ref. 2). Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. Performance of inservice testing discussed in the ASME Code (Ref. 2) (only required at 3 month intervals) satisfies this requirement. The test Frequency in accordance with the Inservice Testing Program results in testing each pump once every 3 months, as required by Reference 2.

When on recirculation, the required differential pressure for the AFW pumps (Ref. 4) when tested in accordance with the Inservice Testing Program is:

Motor Driven Pumps 1514 psid at a nominal flow of 110 gpm Turbine Driven Pump 1616.4 psid at a nominal flow of 130 gpm Wolf Creek - Unit 1 B 3.7.5-8 Revision 90

ENCLOSURE Page 260 of 264 AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.2 (continued)

REQUIREMENTS This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test.

SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.

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 controlled under the Surveillance Frequency Control Program.

This SR includes the requirement to verify that each AFW motor-operated discharge valve limits the flow from the motor driven AFW pump to each steam generator to 320 gpm and that valves in the ESW suction flowpath actuate to the full open position upon receipt of an Auxiliary Feedwater Pump Suction Pressure-Low signal.

SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an AFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

This SR is modified by a Note. The Note indicates that the SR be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test.

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ENCLOSURE Page 261 of 264 AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.5 REQUIREMENTS (continued) This SR verifies that the AFW is properly aligned by verifying the flow paths from the CST to each steam generator prior to entering MODE 2 after more than 30 days in MODE 5 or 6. OPERABILITY of AFW flow paths must be verified before sufficient core heat is generated that would require the operation of the AFW System during a subsequent shutdown.

The Frequency is reasonable, based on engineering judgement and other administrative controls that ensure that flow paths remain OPERABLE. To further ensure AFW System alignment, flow path OPERABILITY is verified following extended outages to determine no misalignment of valves has occurred. This SR ensures that the flow path from the CST to the steam generators is properly aligned.

REFERENCES 1. USAR, Section 10.4.9.

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

3. NRC letter (C. Poslusny to O. Maynard) dated December 16, 1998:

Wolf Creek Generating Station - Technical Specification Bases Change, Auxiliary Feedwater System.

4. Performance Improvement Request 2002-0945.

5. Condition Report 2006-000188.

Wolf Creek - Unit 1 B 3.7.5-10 Revision 90

ENCLOSURE Page 262 of 264 Class 1E Electrical Equipment Air-Conditioning (A/C) System B 3.7.20 BASES APPLICABLE During normal operations each Class 1E electrical equipment A/C train SAFETY ANALYSES maintains the temperature in its associated electrical equipment rooms (continued) at a temperature 90°F. The Class 1E electrical equipment A/C trains are designed in accordance with Seismic Category I requirements.

As confirmed by analysis, the capability of a single Class 1E electrical of electrical equipment during normal conditions, with the implementation of mitigating actions, corresponds to that trains capability to maintain accident conditions.

The Class 1E Electrical Equipment A/C System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Two independent Class 1E electrical equipment A/C trains are required to be OPERABLE to ensure adequate cooling to their associated electrical equipment rooms during normal operation. Each Class 1E electrical equipment A/C train is considered to be OPERABLE when the individual components necessary to maintain associated electrical equipment room temperatures within acceptable limits are OPERABLE. These components include the refrigeration compressors, heat exchangers, cooling coils, ESW flow, fans and associated temperature control instrumentation. In addition, the Class 1E electrical equipment A/C trains must be OPERABLE to the extent that air circulation can be maintained.

APPLICABILITY In MODES 1, 2, 3, and 4, the Class 1E Electrical Equipment A/C System is a normally operating system with both trains in operation. Both trains must be OPERABLE to ensure that the temperature in the protected rooms will not exceed equipment design limits.

Although the LCO for the Class 1E Electrical Equipment A/C System is not applicable in MODES 5 and 6, the capability of the Class 1E Electrical Equipment A/C System to perform its necessary related support functions may be required for OPERABILITY of supported systems.

Wolf Creek - Unit 1 B 3.7.20-2 Revision 90

ENCLOSURE Page 263 of 264 Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES BACKGROUND The capacity and controls associated with the lube oil system are (continued) sufficient to ensure a minimum of 7 days of operation. Refer to the table below:

Approximate Crankcase Description Volume High level alarm 1215 gallons Dipstick full mark 1200 gallons Automatic Makeup valve isolates 1143 gallons Automatic makeup valve opens 1063 gallons Low level alarm 963 gallons Dipstick add oil mark (10 day supply) 948 gallons 7 day supply 750 gallons 6 day supply 686 gallons Unusable volume in crankcase 300 gallons Each DG has an air start system with adequate capacity for five successive start attempts on the DG without recharging the air start receiver(s). The air start system for each DG has two redundant trains each with a separate air receiver. A single train has adequate capacity for five successive start attempts without recharging; however, a single receiver must be at a higher air pressure than two receivers to support this starting capability.

APPLICABLE The initial conditions of Design Basis Accident (DBA) and transient SAFETY ANALYSES analyses in the USAR, Chapter 6 (Ref. 4), and in the USAR, Chapter 15 (Ref. 5), assume Engineered Safety Feature (ESF) systems are OPERABLE. The DGs are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that fuel, Reactor Coolant System and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems.

Since diesel fuel oil, lube oil, and the air start subsystem support the operation of the standby AC power sources, they satisfy Criterion 3 of the 10 CFR 50.36(c)(2)(ii).

Wolf Creek - Unit 1 B 3.8.3-2 Revision 90

ENCLOSURE Page 264 of 264 RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES LCO c. Indication of reactor coolant temperature.

(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 RCS temperature. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs. An OPERABLE RHR loop must be capable of being realigned to provide an OPERABLE flow path. Management of gas voids is important to RHR System OPERABILITY.

When both RHR loops (or trains) are required to be OPERABLE, the associated Component Cooling Water (CCW) train is required to be OPERABLE. The heat sink for the CCW System is normally provided by the Service Water System or Essential Service Water (ESW) System, as determined by system availability. In MODES 5 and 6, one Diesel Generator (DG) is required to be OPERABLE per LCO 3.8.2, AC Sources - Shutdown. The same ESW train is required to be capable of performing its related support function(s) to support DG OPERABILITY.

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 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation - High Water Level."

Since LCO 3.9.6 contains Required Actions with immediate Completion Times related to the restoration of the degraded decay heat removal function, it is not permitted to enter this LCO from either MODE 5 or from LCO 3.9.5, RHR and Coolant Circulation - High Water Level, unless the requirements of LCO 3.9.6 are met. This precludes diminishing the backup decay heat removal capability when the RHR System is degraded.

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 in accordance with the LCO or until 23 ft of water level is established above the reactor Wolf Creek - Unit 1 B 3.9.6-2 Revision 90