Text

Changes to Technical Specification Bases - Revisions 95 and 96
	ML25058A069
Person / Time
Site:	Wolf Creek
Issue date:	02/27/2025
From:	Hamman D; Wolf Creek
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	000855
	Download: ML25058A069 (1)
	v • d • e

P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Dustin T. Hamman Director Nuclear and Regulatory Affairs February 27, 2025 000855 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 95 and 96 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 95 and 96) under the provisions to TS Section 5.5.14 and a List of Effective Pages. This submittal reflects changes from January 1, 2024, through December 31, 2024.

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

Sincerely, Dustin T. Hamman DTH/jkt

Enclosure:

Wolf Creek Generating Station Changes to the Technical Specification Bases cc:

A. N. Agrawal (NRC), w/e S. S. Lee (NRC), w/e J. D. Monninger (NRC), w/e Senior Resident Inspector (NRC), w/e WC Licensing Correspondence, w/e - RA 25-000855

Enclosure to 000855 Wolf Creek Generating Station Changes to the Technical Specification Bases (63 pages)

TS BASES REVISION: 95 TECHNICAL SPECIFICATION BASES Wolf Creek Generating Station, Unit 1 Summary of Revision 95:

1)

Revised TS Bases pages B 3.1.1-3, B 3.1.3-2, B 3.1.3-6, and B 3.1.6-6, These pages were supposed to have been revised in Revision 84 as part of License Amendment 221 but were inadvertently omitted. Amendment 221 implemented the transition to Westinghouse core design and safety analysis methodologies, as well as adopting the alternate source term in accordance with 10 CFR 50.67.

2)

Revised TS Bases pages B 3.3.1-11, B 3.3.1-12, B 3.3.1-36, B 3.3.1-58, and B 3.3.1-60 as part of implementation of License Amendment 240. Amendment 240 revised Technical Specification Table 3.3.1-1, Reactor Trip System Instrumentation, by removing Function 3.b, Power Range Neutron Flux Rate - High Negative Rate Trip.

3)

Revised TS Bases page B 3.3.1-17 to clarify that the axial flux difference (AFD) penalty in Note 2 of Technical Specification Table 3.3.1-1 is not required and was physically deleted from the Overpower Delta-T setpoint circuit by DCP 012858. (CR 10027080)

4)

Revised TS Bases page B 3.7.16-2 to correct an error that occurred in Revision 23. TS LCO 3.7.16 requires that the fuel storage pool boron concentration shall be 2165 ppm. In Revision 23, the TS Bases page B 3.7.16-2 was inadvertently changed to state The fuel storage pool boron concentration is required to = 2165 ppm.

5)

Revised TS Bases page B 3.8.1-3 to reflect changes being made to the switchyard during Refueling Outage 26. (Change Package 020516)

6)

Revised TS Bases page B 3.8.1-18 to reflect that the frequency of all TS required surveillance testing shall be in accordance with the Surveillance Frequency Control Program. (CR 10029828)

7)

An editorial correction was made on TS Bases page B 3.8.5-2 to reflect accurate paragraph lettering.

8)

TS Bases page B 3.8.6-5 was inadvertently removed and had an extra copy of page B 3.8.7-3 inserted in its place in Revision 94. This change replaces the previous page B 3.8.6-5 as it was prior to Revision 94.

9)

Administrative correction of several lines in the List of Effective Pages that referred to incorrect revision numbers. These changes are not marked with change bars and no changes to the actual pages are required.

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Wolf Creek - Unit 1 i Revision 95 TAB - Title Page Technical Specification Cover Page Title Page TAB - Table of Contents i

81 DRR 19-1027 10/28/19 ii 81 DRR 19-1027 10/28/19 iii 81 DRR 19-1027 10/28/19 TAB - B 2.0 SAFETY LIMITS (SLs)

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Wolf Creek - Unit 1 ii Revision 95 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)

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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

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Wolf Creek - Unit 1 iii Revision 95 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

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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 95 N/A 4/25/24 B 3.3.1-12 95 N/A 4/25/24 B 3.3.1-13 95 N/A 4/25/24 B 3.3.1-14 95 N/A 4/25/24 B 3.3.1-15 95 N/A 4/25/24 B 3.3.1-16 95 N/A 4/25/24 B 3.3.1-17 95 N/A 4/25/24 B 3.3.1-18 95 N/A 4/25/24 B 3.3.1-19 95 N/A 4/25/24 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

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Wolf Creek - Unit 1 iv Revision 95 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 95 N/A 4/25/24 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 84 DRR 20-0400 08/18/20 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 95 N/A 4/25/24 B 3.3.1-59 89 DRR 21-0966 7/7/21 B 3.3.1-60 95 N/A 4/25/24 B 3.3.1-61 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

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Wolf Creek - Unit 1 v Revision 95 TAB - B 3.3 INSTRUMENTATION (continued)

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

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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

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Wolf Creek - Unit 1 vi Revision 95 TAB - B 3.3 INSTRUMENTATION (continued)

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

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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

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Wolf Creek - Unit 1 vii Revision 95 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

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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 89 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 90 DRR 21-1229 11/3/21 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 90 DRR 21-1229 11/3/21 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 B 3.4.12-10 0

Amend. No. 123 12/18/99

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Wolf Creek - Unit 1 viii Revision 95 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

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 94 N/A 10/6/23 B 3.4.13-3 94 N/A 10/6/23 B 3.4.13-4 94 N/A 10/6/23 B 3.4.13-5 94 N/A 10/6/23 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 82 DRR 20-0077 1/29/20 B 3.4.16-2 84 DRR 20-0400 08/18/20 B 3.4.16-3 82 DRR 20-0077 1/29/20 B 3.4.16-4 92 DRR 22-0767 11/3/22 B 3.4.16-5 92 DRR 22-0767 11/3/22 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 29 DRR 06-1984 10/17/06 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 B 3.5.2-5 72 DRR 15-1918 10/26/15

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Wolf Creek - Unit 1 ix Revision 95 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

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 90 DRR 21-1229 11/3/21 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 90 DRR 21-1229 11/3/21 B 3.6.3-8 36 DRR 08-0255 3/11/08 B 3.6.3-9 90 DRR 21-1299 11/3/21 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 B 3.6.3-17 36 DRR 08-0255 3/11/08

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REVISION NO. (2)

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Wolf Creek - Unit 1 x Revision 95 TAB - B 3.6 CONTAINMENT SYSTEMS (continued)

B 3.6.3-18 36 DRR 08-0255 3/11/08 B 3.6.3-19 82 DRR 20-0077 1/29/20 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 82 DRR 20-0077 1/29/20 B 3.7.2-3 82 DRR 20-0077 1/29/20 B 3.7.2-4 81 DRR 19-1027 10/28/19 B 3.7.2-5 82 DRR 20-0077 1/29/20 B 3.7.2-6 82 DRR 20-0077 1/29/20 B 3.7.2-7 82 DRR 20-0077 1/29/20 B 3.7.2-8 82 DRR 20-0077 1/29/20 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 B 3.7.3-7 37 DRR 08-0503 4/8/08

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REVISION NO. (2)

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Wolf Creek - Unit 1 xi Revision 95 TAB - B 3.7 PLANT SYSTEMS (continued)

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 90 DRR 21-1229 11/3/21 B 3.7.5-8 90 DRR 21-1229 11/3/21 B 3.7.5-9 90 DRR 21-1229 11/3/21 B 3.7.5-10 90 DRR 21-1229 11/3/21 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 77 DRR 17-1001 6/22/17 B 3.7.7-3 92 DRR 22-0767 11/3/22 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 B 3.7.13-2 81 DRR 19-1027 10/28/19

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REVISION NO. (2)

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Wolf Creek - Unit 1 xii Revision 95 TAB - B 3.7 PLANT SYSTEMS (continued)

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 95 N/A 4/25/24 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 B 3.7.20-1 B 3.7.20-2 B 3.7.20-3 B 3.7.20-4 B 3.7.20-5 89 79 90 85 89 89 DRR 21-0966 DRR 18-1579 DRR 21-1229 DRR 20-0988 DRR 21-0966 DRR 21-0966 7/7/21 10/22/18 11/3/21 10/24/20 7/7/21 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 93 DRR 22-1363 2/2/23 B 3.8.1-3 B 3.8.1-4 95 93 N/A DRR 22-1363 4/25/24 2/2/23 B 3.8.1-5 93 DRR 22-1363 2/2/23 B 3.8.1-6 93 DRR 22-1363 2/2/23 B 3.8.1-7 93 DRR 22-1363 2/2/23 B 3.8.1-8 93 DRR 22-1363 2/2/23 B 3.8.1-9 93 DRR 22-1363 2/2/23 B 3.8.1-10 93 DRR 22-1363 2/2/23 B 3.8.1-11 93 DRR 22-1363 2/2/23 B 3.8.1-12 93 DRR 22-1363 2/2/23 B 3.8.1-13 93 DRR 22-1363 2/2/23 B 3.8.1-14 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23 B 3.8.1-18 95 N/A 4/25/24

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Wolf Creek - Unit 1 xiii Revision 95 TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

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 93 DRR 22-1363 2/2/23 B 3.8.1-22 89 DRR 21-0966 7/7/21 B 3.8.1-23 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23 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 92 DRR 22-0767 11/3/22 B 3.8.2-4 92 DRR 22-0767 11/3/22 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 90 DRR 21-1229 11/3/21 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 93 DRR 22-1363 2/2/23 B 3.8.4-4 0

Amend. No. 123 12/18/99 B 3.8.4-5 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23 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 95 N/A 4/25/24 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 B 3.8.6-5 0

Amend. No. 123 12/18/99

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Wolf Creek - Unit 1 xiv Revision 95 TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

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 93 DRR 22-1363 2/2/23 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 B 3.9.6-2 90 DRR 21-1229 11/3/21

LIST OF EFFECTIVE PAGES - TECHNICAL SPECIFICATION BASES PAGE (1)

REVISION NO. (2)

CHANGE DOCUMENT (3)

DATE EFFECTIVE/

IMPLEMENTED (4)

Wolf Creek - Unit 1 xv Revision 95 TAB - B 3.9 REFUELING OPERATIONS (continued)

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. As of Revision 94, the TS Bases is no longer controlled in PMAC.

Therefore, starting with Revision 94, the DRR number will be N/A.

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

SDM B 3.1.1 Wolf Creek - Unit 1 B 3.1.1-3 Revision 95 BASES APPLICABLE with respect to potential fuel damage is a guillotine break of a main SAFETY ANALYSES steamline inside containment initiated at the end of core life with RCS Tavg (continued) equal to 557°F. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.

In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.

Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.

The startup of an inactive RCP is administratively precluded in MODES 1 and 2. In MODE 3, the startup of an inactive RCP can not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition.

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

SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-2 Revision 95 BASES BACKGROUND The SRs for measurement of the MTC at the beginning and near the end (continued) of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.

APPLICABLE The acceptance criteria for the specified MTC are:

SAFETY ANALYSES

a.

The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2); and

b.

The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

The USAR, Chapter 15 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).

The consequences of accidents that cause core overheating must be evaluated when the MTC is positive (part-power conditions) or zero (full-power conditions). Such accidents include the rod withdrawal transients from either subcritical or at-power conditions, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.

In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2).

MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted at conditions when the RCS boron concentration reaches a boron concentration equivalent to 300 ppm at an equilibrium, all rods out, RTP condition. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-6 Revision 95 BASES SURVEILLANCE SR 3.1.3.2 (continued)

REQUIREMENTS

2.

If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC.

Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit.

3.

The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is less negative than the 60 ppm Surveillance limit, the EOC limit will not be exceeded because of the gradual manner in which MTC changes with core burnup REFERENCES

1.

10 CFR 50, Appendix A, GDC 11.

2.

USAR, Chapter 15.

3.

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

Control Bank Insertion Limits B 3.1.6 Wolf Creek - Unit 1 B 3.1.6-6 Revision 95 BASES REFERENCES

1.

10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.

2.

10 CFR 50.46.

3.

USAR, Chapter 15.

4.

USAR, Section 4.3.1.5.

5.

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

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-11 Revision 95 BASES APPLICABLE

b.

Power Range Neutron Flux - Low (continued)

SAFETY ANALYSES, LCO, and In addition, in MODE 3 (with any RCS cold leg temperature APPLICABILITY

< 500° F, or the RCS sufficiently borated, or the RCCA bank withdrawal event precluded per the specified conditions of footnote (i) in Table 3.3.1-1), 4, 5, or 6, the Power Range Neutron Flux - Low trip Function does not have to be OPERABLE because the reactor is shut down and the NIS power range detectors cannot accurately detect neutron levels in this range. Other RTS trip Functions and administrative controls provide protection against positive reactivity excursions in these MODES and specified conditions in the Applicability.

3.

Power Range Neutron Flux Rate - High Positive Rate The Power Range Neutron Flux Rate trip uses the same channels as discussed for Function 2 above.

The Power Range Neutron Flux - High Positive Rate trip Function ensures that protection is provided against rapid increases in neutron flux that are characteristic of an RCCA drive rod housing rupture and the accompanying ejection of the RCCA. This Function compliments the Power Range Neutron Flux - High and Low Setpoint trip Functions to ensure that the criteria are met for a rod ejection from the power range. This Function also provides protection for the rod withdrawal at power event.

The LCO requires all four of the Power Range Neutron Flux - High Positive Rate channels to be OPERABLE.

In MODE 1 or 2, when there is a potential to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron Flux - High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the Power Range Neutron Flux - High Positive Rate trip Function does not have to be OPERABLE because other RTS trip Functions and administrative controls will provide protection against positive reactivity excursions.

4.

Intermediate Range Neutron Flux The Intermediate Range Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition. This trip Function provides redundant protection to the Power Range Neutron Flux - Low Setpoint trip Function. The NIS intermediate range

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-12 Revision 95 BASES APPLICABLE

4.

Intermediate Range Neutron Flux (continued)

SAFETY ANALYSES, LCO, and detectors are located external to the reactor vessel and measure APPLICABILITY neutrons leaking from the core. The NIS intermediate range detectors do not provide any input to control systems. Note that this Function also provides a signal to prevent automatic and manual rod withdrawal prior to initiating a reactor trip. Limiting further rod withdrawal may terminate the transient and eliminate the need to trip the reactor.

The LCO requires two channels of Intermediate Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function.

The Trip Setpoint is 25% RTP.

Because this trip Function is important only during startup, there is generally no need to disable channels for testing while the Function is required to be OPERABLE. Therefore, a third channel is unnecessary.

In MODE 1 below the P-10 setpoint, and in MODE 2, above the P6 setpoint when there is a potential for an uncontrolled RCCA bank rod withdrawal accident during reactor startup, the Intermediate Range Neutron Flux trip must be OPERABLE.

Above the P-10 setpoint, the Power Range Neutron Flux - High Setpoint trip and the Power Range Neutron Flux - High Positive Rate trip provide core protection for an uncontrolled RCCA bank withdrawal accident. In MODE 2 (below the P-6 setpoint), the Source Range Neutron Flux trip Function provides core protection for reactivity accidents. In MODE 3, 4, 5, or 6, the Intermediate Range Neutron Flux trip Function does not have to be OPERABLE.

In MODE 3 with all RCS cold leg temperatures 500° F, and the RCS boron concentration less than or equal to the ARO critical boron concentration, the Rod Control System capable of rod withdrawal or one or more rods not fully inserted, the Power Range Neutron Flux - Low trip Function provides protection for an uncontrolled RCCA bank withdrawal or control rod ejection event from low power or subcritical conditions.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-13 Revision 95 BASES APPLICABLE

4.

Intermediate Range Neutron Flux (continued)

SAFETY ANALYSES, LCO, and With the Rod Control System capable of rod withdrawal in MODE APPLICABILITY 3 with any RCS cold leg temperature < 500° F, in MODE 4, or MODE 5, LCO 3.1.9, RCS Boron Limitations < 500o, requires that the RCS boron concentration be greater than the ARO critical boron concentration to ensure that sufficient SDM is available if an uncontrolled RCCA bank withdrawal event were to occur. In MODE 6, the Rod Control System is incapable of rod withdrawal and the core has a required increased SDM. Also, the NIS intermediate range detectors cannot adequately detect neutron levels present during lower temperatures.

5.

Source Range Neutron Flux The LCO requirement for the Source Range Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition. This trip Function provides redundant protection to the Power Range Neutron Flux - Low trip Function.

In MODES 3, 4, and 5, administrative controls also prevent the uncontrolled withdrawal of rods. The NIS source range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS source range detectors do not provide any inputs to control systems. The source range trip is the only RTS automatic protection function required in MODES 3 (with any RCS cold leg temperature < 500o F), 4, and 5 with the Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

In MODE 3 with all RCS cold leg temperatures 500° F, and the RCS boron concentration less than or equal to the ARO critical boron concentration, the Rod Control System capable of rod withdrawal or one or more rods not fully inserted, the Power Range Neutron Flux - Low trip Function provides protection for an uncontrolled RCCA bank withdrawal or control rod ejection event from low power or subcritical conditions.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-14 Revision 95 BASES APPLICABLE

5.

Source Range Neutron Flux (continued)

SAFETY ANALYSES, LCO, and With the Rod Control System capable of rod withdrawal in MODE APPLICABILITY 3 with any RCS cold leg temperature < 500° F, in MODE 4, or MODE 5, LCO 3.1.9, RCS Boron Limitations < 500° F, requires that the RCS boron concentration be greater than the ARO critical boron concentration to ensure that sufficient SDM is available if an uncontrolled RCCA bank withdrawal event were to occur. The safety analyses do not take explicit credit for the Source Range Neutron Flux trip Function as a primary trip to mitigate an uncontrolled RCCA bank withdrawal event or control rod ejection occurring from low power or subcritical conditions since this trip Function is not tested for its response time under SR 3.3.1.16.

LCO 3.1.9, RCS Boron Limitations < 500° F, assures that sufficient SDM is available if an uncontrolled RCCA bank withdrawal were to occur while the plant is operating within that LCOs Applicability and specified conditions.

The LCO requires two channels of Source Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function. The Trip Setpoint is 1.0 E5 cps. The outputs of the Function to RTS logic are not required OPERABLE in MODE 6 or when all rods are fully inserted and the Rod Control System is incapable of rod withdrawal.

The Source Range Neutron Flux Function provides protection for control rod withdrawal from subcritical and control rod ejection events.

In MODE 2, when below the P-6 setpoint, the Source Range Neutron Flux trip must be OPERABLE. Above the P-6 setpoint, the Intermediate Range Neutron Flux trip and the Power Range Neutron Flux - Low trip will provide core protection for reactivity accidents. Above the P-6 setpoint, the NIS source range detectors are de-energized.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-15 Revision 95 BASES APPLICABLE

5.

Source Range Neutron Flux (continued)

SAFETY ANALYSES, LCO, and In MODE 3, 4, or 5 with the Rod Control System capable of rod APPLICABILITY withdrawal or one or more rods not fully inserted, the Source Range Neutron Flux trip Function must also be OPERABLE. If the Rod Control System is capable of rod withdrawal, the Source Range Neutron Flux trip must be OPERABLE to provide core protection against a rod withdrawal accident. If the Rod Control System is not capable of rod withdrawal with rods fully inserted in MODES 3, 4, and 5, the source range detectors are not required to trip the reactor. However, their monitoring Function must be OPERABLE to monitor core neutron levels and provide indication of reactivity changes that may occur. The requirements for the NIS source range detectors in MODE 6 are addressed in LCO 3.9.3, "Nuclear Instrumentation."

6.

Overtemperature T The Overtemperature T trip Function is provided to ensure that the design limit DNBR is met. This trip Function also limits the range over which the Overpower T trip Function must provide protection. The inputs to the Overtemperature T trip include pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop T assuming full reactor coolant flow. Protection from violating the DNBR limit is assured for those transients that are slow with respect to delays from the core to the measurement system. The Overtemperature T trip Function uses each loop's T as a measure of reactor power and is compared with a setpoint that is automatically varied with the following parameters:

reactor coolant average temperature - the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; pressurizer pressure - the Trip Setpoint is varied to correct for changes in system pressure; and axial power distribution f(I) - the Trip Setpoint is varied to account for imbalances in the axial power distribution as detected by the NIS upper and lower power range detectors. If axial peaks are greater than the design limits, as indicated by the difference between the upper and lower NIS power range detectors, the Trip Setpoint is reduced.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-16 Revision 95 BASES APPLICABLE

6.

Overtemperature T (continued)

SAFETY ANALYSES, LCO, and Dynamic compensation is included for system piping delays from APPLICABILITY the core to the temperature measurement system.

The Overtemperature T trip Function is calculated for each loop as described in Note 1 of Table 3.3.1-1. Trip occurs if Overtemperature T is indicated in two loops. The pressure and temperature signals are used for other control functions; thus, the actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Trip Setpoint. A turbine runback will reduce turbine power and reactor power. A reduction in reactor power will normally alleviate the Overtemperature T condition and may prevent a reactor trip.

The LCO requires all four channels of the Overtemperature T trip Function to be OPERABLE. Note that the Overtemperature T Function receives input from channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

In MODE 1 or 2, the Overtemperature T trip must be OPERABLE to prevent DNB. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about DNB.

7.

Overpower T The Overpower T trip Function ensures that protection is provided to ensure the integrity of the fuel (i.e., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions. This trip Function also limits the required range of the Overtemperature T trip Function and provides a backup to the Power Range Neutron Flux - High Setpoint trip. The Overpower T trip Function ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-17 Revision 95 BASES APPLICABLE

7.

Overpower T (continued)

SAFETY ANALYSES, LCO, and The Overpower T trip also provides protection to mitigate the APPLICABILITY consequences of a small steamline break, as reported in Ref. 8, and a steamline break with coincident control rod withdrawal (Ref.

9). It uses the T of each loop as a measure of reactor power with a setpoint that is automatically varied with the following parameters:

reactor coolant average temperature - the Trip Setpoint is varied to correct for changes in coolant density and specific heat capacity with changes in coolant temperature; and rate of change of reactor coolant average temperature -

including dynamic compensation for the delays between the core and the temperature measurement system.

The Overpower T trip Function is calculated for each loop as per Note 2 of Table 3.3.1-1. Trip occurs if Overpower T is indicated in two loops. The temperature signals are used for other control functions. The actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation and a single failure in the remaining channels providing the protection function actuation. Note that this Function also provides a signal to generate a turbine runback prior to reaching the Trip Setpoint. A turbine runback will reduce turbine power and reactor power. A reduction in power will normally alleviate the Overpower T condition and may prevent a reactor trip. The axial flux difference (AFD) penalty function generator cards were removed from the Overpower T setpoint circuits per DCP 012858. As such, AFD is not an input to the Overpower T setpoint and malfunctions in the Power Range NI channel cannot propagate to the Overpower T setpoint.

The LCO requires four channels of the Overpower T trip Function to be OPERABLE. Note that the Overpower T trip Function receives input from channels shared with other RTS Functions. Failures that affect multiple Functions require entry into the Conditions applicable to all affected Functions.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-18 Revision 95 BASES APPLICABLE

7.

Overpower T (continued)

SAFETY ANALYSES LCO, and In MODE 1 or 2, the Overpower T trip Function must be APPLICABILITY OPERABLE. These are the only times that enough heat is generated in the fuel to be concerned about the heat generation rates and overheating of the fuel. In MODE 3, 4, 5, or 6, this trip Function does not have to be OPERABLE because the reactor is not operating and there is insufficient heat production to be concerned about fuel overheating and fuel damage.

8.

Pressurizer Pressure Four Pressurizer Pressure channels provide input to the Pressurizer Pressure - High and - Low trips and the Overtemperature T trip. The Pressurizer Pressure channels are also used to provide input to the Pressurizer Pressure Control System; thus, the actuation logic must be able to withstand an input failure to the control system, which may then require the protection function actuation, and a single failure in the other channels providing the protection function actuation.

a.

Pressurizer Pressure - Low The Pressurizer Pressure - Low trip Function ensures that protection is provided against violating the DNBR limit due to low pressure.

The LCO requires four channels of Pressurizer Pressure-Low to be OPERABLE. The Trip Setpoint is 1940 psig.

In MODE 1, when DNB is a major concern, the Pressurizer Pressure - Low trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock (NIS power range P-10 or turbine impulse pressure greater than approximately 10% of full power equivalent (P-13). On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, there is insufficient heat production to generate DNB conditions.

b.

Pressurizer Pressure - High The Pressurizer Pressure - High trip Function ensures that protection is provided against overpressurizing the RCS.

This trip Function operates in conjunction with the pressurizer relief and safety valves to prevent RCS overpressure conditions.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-19 Revision 95 BASES APPLICABLE

b.

Pressurizer Pressure - High (continued)

SAFETY ANALYSES, LCO, and The LCO requires four channels of the Pressurizer APPLICABILITY Pressure - High to be OPERABLE. The Trip Setpoint is 2385 psig.

The Pressurizer Pressure - High Allowable Value is selected to be below the pressurizer safety valve actuation pressure and above the power operated relief valve (PORV) setting. This setting minimizes challenges to safety valves while avoiding unnecessary reactor trip for those pressure increases that can be controlled by the PORVs.

In MODE 1 or 2, the Pressurizer Pressure - High trip must be OPERABLE to help prevent RCS overpressurization and minimize challenges to the relief and safety valves. In MODE 3, 4, 5, or 6, the Pressurizer Pressure - High trip Function does not have to be OPERABLE because transients that could cause an overpressure condition will be slow to occur. Therefore, the operator will have sufficient time to evaluate unit conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure protection when the temperature of one or more RCS loops is below 368°F.

9.

Pressurizer Water Level - High The Pressurizer Water Level - High trip Function provides a backup signal for the Pressurizer Pressure - High trip and also provides protection against water relief through the pressurizer safety valves. These valves are designed to pass steam in order to achieve their design energy removal rate. A reactor trip is actuated prior to the pressurizer becoming water solid. The LCO requires three channels of Pressurizer Water Level - High to be OPERABLE. The Trip Setpoint is 92% of instrument span. The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is not required to address control/protection interaction concerns. The level channels do not actuate the safety valves, and the high pressure reactor trip is set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip.

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-36 Revision 95 BASES ACTIONS E.1 and E.2 (continued)

Overtemperature T; Overpower T; Power Range Neutron Flux - High Positive Rate; Pressurizer Pressure - High; and SG Water Level - Low Low.

A known inoperable channel must be placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Placing the channel in the tripped condition results in a partial trip condition requiring only one-out-of-three logic for actuation of the two-out-of-four trip logic. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed to place the inoperable channel in the tripped condition is justified in Reference 12.

If the inoperable channel cannot be placed in the trip condition within the specified Completion Time, the unit must be placed in a MODE where these Functions are not required OPERABLE. An additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed to place the unit in MODE 3. Six hours is a reasonable time, based on operating experience, to place the unit in MODE 3 from full power in an orderly manner and without challenging unit systems.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while performing routine surveillance testing of the other channels. The 12-hour time limit is justified in Reference 12.

F.1 and F.2 Condition F applies to the Intermediate Range Neutron Flux trip when THERMAL POWER is above the P-6 setpoint and below the P-10 setpoint and one channel is inoperable. Above the P-6 setpoint and below the P-10 setpoint, the NIS intermediate range detector performs the monitoring Functions. If THERMAL POWER is greater than the P-6 setpoint but less than the P-10 setpoint, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed to reduce THERMAL POWER below the P-6 setpoint or to increase THERMAL POWER above the P-10 setpoint. The NIS Intermediate Range Neutron Flux channels must be OPERABLE when the power level is above the capability of the source range, P-6, and below the capability of the power range, P-10. If THERMAL POWER is greater than the P-10 setpoint, the NIS power range detectors perform the monitoring and protection

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-58 Revision 95 TABLE B 3.3.1-1 (Page 1 of 2)

FUNCTION TRIP SETPOINT(a)

1.

Manual Reactor Trip

2. Power Range Neutron Flux

a. High

b. Low

3. Power Range Neutron Flux High Positive Rate

4. Intermediate Range Neutron Flux

5. Source Range Neutron Flux

6. Overtemperature ¨T

7. Overpower ¨T

8. Pressurizer Pressure

a. Low

b. High

9. Pressurizer Water level - High

10. Reactor Coolant Flow - Low

11. Not Used

12. Undervoltage RCPs

13. Underfrequency RCPs

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

15. Not Used

16. Turbine Trip

a. Low Fluid Oil Pressure

b. Turbine Stop Valve Closure NA 109% of RTP 25% of RTP 4% of RTP with a time constant 2 seconds 25% of RTP 105 cps See Table 3.3.1-1 Note 1 See Table 3.3.1-1 Note 2 1940 psig 2385 psig 92% of instrument span 89.9% of Normalized Flow 10578 Vac 57.15 Hz 23.5% of narrow range instrument span 590.00 psig 1% open

RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-60 Revision 95 TABLE B 3.3.1-2 (Page 1 of 2)

FUNCTIONAL UNIT RESPONSE TIME

1.

Manual Reactor Trip

2.

Power Range Neutron Flux

a. High

b. Low

3.

Power Range Neutron Flux High Positive Rate

4.

Intermediate Range Neutron Flux

5.

Source Range Neutron Flux

6.

Overtemperature ¨T

7.

Overpower ¨T

8.

Pressurizer Pressure

a. Low

b. High

9.

Pressurizer Water Level - High

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 N.A.

0.5 second(1) 0.5 second(1) 0.5 second(1).

N.A.

6.0 seconds(1) 6.0 seconds(1) 2.0 seconds 1.0 second N.A.

1.0 second 1.0 second 1.5 seconds 0.6 second 2.0 seconds (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.

Fuel Storage Pool Boron Concentration B 3.7.16 Wolf Creek - Unit 1 B 3.7.16-2 Revision 95 BASES BACKGROUND each assembly in accordance with LCO 3.7.17, Spent Fuel Assembly (continued)

Storage. Prior to movement of an assembly, it is necessary to perform SR 3.7.16.1.

APPLICABLE Accidents can be postulated that could increase the reactivity of the fuel SAFETY ANALYSES storage pool which are unacceptable with unborated water in the fuel storage pool. Thus, for these accident occurrences, the presence of soluble boron in the storage pool maintains subcriticality with a Keff of 0.95 or less. The postulated accidents are basically of two types. Multiple fuel assemblies could be incorrectly transferred to non-Region 1 locations (e.g., unirradiated fuel assemblies or insufficiently depleted fuel assemblies). The second type of postulated accidents is associated with a fuel assembly which is dropped adjacent to the fully loaded storage rack. The negative reactivity effect of the soluble boron compensates for the increased reactivity caused by either one of the two postulated accident scenarios. The accident analyses is provided in the USAR, Appendix 9.1A (Ref. 1).

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The fuel storage pool boron concentration is required to be 2165 ppm.

The fuel storage pool consists of the spent fuel pool and cask loading pool (with racks installed). The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 1.

This concentration of dissolved boron is the minimum required concentration for non-inventoried fuel assembly storage and movement within the fuel storage pool.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the fuel storage pool, until a complete fuel storage pool verification has been performed following the last movement of fuel assemblies in the fuel storage pool.

This verification shall consist of a confirmation of the fuel assembly serial number of every assembly moved in the fuel storage pool since the last verification. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies.

With no further fuel assembly movements in progress, there is no potential for misloaded fuel assemblies or a dropped fuel assembly.

AC Sources - Operating B 3.8.1 Wolf Creek - Unit 1 B 3.8.1-3 Revision 95 BASES APPLICABLE Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS);

SAFETY ANALYSES and Section 3.6, Containment Systems.

(continued)

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

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 #9/#11 transformers, fed from the 345 kV offsite transmission network, feeding power through breaker 13-49 to the ESF transformer XNB01, which, in turn powers the NB01 bus through its normal feeder breaker or the NB02 bus through its alternate feeder breaker. Transformer XNB01 may also be powered from the #8/#10 transformers, fed from the 345 kV offsite transmission network, feeding power through breaker 13-50.

Another offsite circuit consists of the startup transformer feeding through breaker PA0201 powering the ESF transformer XNB02, which, in turn powers the NB02 bus through its normal feeder breaker or the NB01 bus through its alternate feeder breaker.

AC Sources - Operating B 3.8.1 Wolf Creek - Unit 1 B 3.8.1-18 Revision 95 BASES ACTIONS I.1 (continued)

Condition I corresponds to a level of degradation in which all redundancy in the AC electrical power supplies has been lost. At this severely degraded level, any further losses in the AC electrical power system will cause a loss of function. Therefore, no additional time is justified for continued operation. The unit is required by LCO 3.0.3 to commence a controlled shutdown.

SURVEILLANCE The AC sources are designed to permit inspection and testing of all REQUIREMENTS important areas and features, especially those that have a standby function, in accordance with 10 CFR 50, Appendix A, GDC 18 (Ref. 8).

Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions). The SRs for demonstrating the OPERABILITY of the DGs are in accordance with the recommendations of Regulatory Guide 1.9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), and Regulatory Guide 1.137 (Ref. 10), as addressed in the USAR. Surveillance testing frequency will be in accordance with the Surveillance Frequency Control Program.

Where the SRs discussed herein specify voltage and frequency tolerances, the following is applicable. This minimum steady state output voltage of 3950 V is 95% of the nominal 4160 V output voltage. This value, which is 210 V above the minimum utilization voltage specified in ANSI C84.1 (Ref. 11), allows for voltage drop to the terminals of 4000 V motors whose minimum operating voltage is specified as 90% or 3600 V.

It also allows for voltage drops to motors and other equipment down through the 120 V level. This value provides for the OPERABILITY of required loads as shown by load flow calculations in support of NRC Branch Technical Position PSB-1. These calculations have demonstrated that no end use loads will be adversely affected from sustained operation above the degraded voltage allowable value as specified in SR 3.3.5.3.

The 3950 V is above the calculated allowable value. The specified maximum steady state output voltage of 4320 V ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages. The specified minimum and maximum frequencies of the DG are 59.4 Hz and 60.6 Hz.

SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that distribution buses and loads are connected to their preferred power source, and that appropriate

DC Sources - Shutdown B 3.8.5 Wolf Creek - Unit 1 B 3.8.5-2 Revision 95 BASES APPLICABLE design requirements during shutdown conditions are allowed by the LCO SAFETY ANALYSES for required systems.

(continued)

During MODES 1, 2, 3, and 4, various deviations from the analysis assumptions and design requirements are allowed within the Required Actions. This allowance is in recognition that certain testing and maintenance activities must be conducted provided an acceptable level of risk is not exceeded. During MODES 5 and 6, performance of a significant number of required testing and maintenance activities is also required. In MODES 5 and 6, the activities are generally planned and administratively controlled. Relaxations from MODE 1, 2, 3, and 4 LCO requirements are acceptable during shutdown modes based on:

a.

The fact that time in an outage is limited. This is a risk prudent goal as well as a utility economic consideration.

b.

Requiring appropriate compensatory measures for certain conditions. These may include administrative controls, reliance on systems that do not necessarily meet typical design requirements applied to systems credited in operating MODE analyses, or both.

c.

Prudent utility consideration of the risk associated with multiple activities that could affect multiple systems.

d.

Maintaining, to the extent practical, the ability to perform required functions (even if not meeting MODE 1, 2, 3, and 4 OPERABILITY requirements) with systems assumed to function during an event.

In addition to the requirements established by the Technical Specifications, the plant staff must also manage shutdown tasks and electrical support to maintain risk at an acceptably low value.

As required by the Technical Specifications, one train of the required equipment during shutdown conditions is supported by one train of AC and DC power and distribution. The availability of additional equipment, both redundant equipment as required by the Technical Specifications and equipment not required by the specifications, contributes to risk reduction and this equipment should be supported by reliable electrical power systems. Typically the Class 1E power sources and distribution systems of the unit are used to power this equipment because these power and distribution systems are available and reliable. When portions of the Class 1E power or distribution systems are not available (usually as a result of maintenance or modifications), other reliable power sources or distribution are used to provide the needed electrical support. The plant staff assesses these alternate power sources and distribution systems to

TS BASES REVISION: 96 TECHNICAL SPECIFICATION BASES Wolf Creek Generating Station, Unit 1 Summary of Revision 96:

1)

Revised TS Bases pages B 3.1.3-1 through B 3.1.3-6 based on Westinghouse PA-ASC-1588, Moderator Temperature Coefficient (MTC) Tech Spec Surveillance Requirements Verification vs.

Measurement. This change will allow MTC to be verified at both BOC and EOC using pre-established criteria, rather than having to do direct measurements. (CR 00133993)

2)

Revised TS Bases pages B 3.1.4-8, B 3.1.6-5, B 3.1.8-5, B 3.3.5-4, B 3.3.8-1, and B 3.5.4-5 to correct the Revision numbers shown at the bottom of the pages. (CR 10032609)

3)

Revised TS Bases page B 3.6.1-4 to clarify the as-left acceptance criteria for SR 3.6.1.1 uses maximum pathway leakage summation only in an outage when only Type B and C testing is performed.

4)

Revised TS Bases page B 3.7.13-5 to change Reference 10 to Reference 8 discussed in the last sentence on the page. Revision 89 of the TS Bases revised and renumbered the reference documents such that Reg Guide 1.52 is now listed as Reference 8. (CR 10034570)

5)

Revised TS Bases page B 3.8.1-4 LCO section to describe that a circuit may be connected to more than one ESF bus and not violate separation criteria, but the offsite circuit not connected to an ESF bus in this alignment must be considered inoperable. (CR 10028226)

6)

Revised TS Bases page B 3.8.1-29 SR 3.8.1.16 to add clarification regarding testing for the alternate power source to be able to verify the ability to manually synchronize and transfer loads.

(CR 10028226)

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Wolf Creek - Unit 1 vi Revision 96 TAB - B 3.3 INSTRUMENTATION (continued)

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 B 3.4.3-7 0

Amend. No. 123 12/18/99

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Wolf Creek - Unit 1 vii Revision 96 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

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 89 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 90 DRR 21-1229 11/3/21 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 90 DRR 21-1229 11/3/21 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 B 3.4.12-10 0

Amend. No. 123 12/18/99

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Wolf Creek - Unit 1 viii Revision 96 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

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 94 N/A 10/6/23 B 3.4.13-3 94 N/A 10/6/23 B 3.4.13-4 94 N/A 10/6/23 B 3.4.13-5 94 N/A 10/6/23 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 82 DRR 20-0077 1/29/20 B 3.4.16-2 84 DRR 20-0400 08/18/20 B 3.4.16-3 82 DRR 20-0077 1/29/20 B 3.4.16-4 92 DRR 22-0767 11/3/22 B 3.4.16-5 92 DRR 22-0767 11/3/22 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 29 DRR 06-1984 10/17/06 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 B 3.5.2-5 72 DRR 15-1918 10/26/15

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Wolf Creek - Unit 1 ix Revision 96 TAB - B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

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 96 N/A 12/11/2024 B 3.6.1-5 96 N/A 12/11/2024 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 90 DRR 21-1229 11/3/21 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 90 DRR 21-1229 11/3/21 B 3.6.3-8 36 DRR 08-0255 3/11/08 B 3.6.3-9 90 DRR 21-1299 11/3/21 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 (New)

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REVISION NO. (2)

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Wolf Creek - Unit 1 x

Revision 96 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 82 DRR 20-0077 1/29/20 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 82 DRR 20-0077 1/29/20 B 3.7.2-3 82 DRR 20-0077 1/29/20 B 3.7.2-4 81 DRR 19-1027 10/28/19 B 3.7.2-5 82 DRR 20-0077 1/29/20 B 3.7.2-6 82 DRR 20-0077 1/29/20 B 3.7.2-7 82 DRR 20-0077 1/29/20 B 3.7.2-8 82 DRR 20-0077 1/29/20 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

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Wolf Creek - Unit 1 xi Revision 96 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 90 DRR 21-1229 11/3/21 B 3.7.5-8 90 DRR 21-1229 11/3/21 B 3.7.5-9 90 DRR 21-1229 11/3/21 B 3.7.5-10 90 DRR 21-1229 11/3/21 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 77 DRR 17-1001 6/22/17 B 3.7.7-3 92 DRR 22-0767 11/3/22 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

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Wolf Creek - Unit 1 xii Revision 96 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 96 N/A 12/11/2024 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 95 N/A 4/25/24 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 B 3.7.20-1 B 3.7.20-2 B 3.7.20-3 B 3.7.20-4 B 3.7.20-5 89 79 90 85 89 89 DRR 21-0966 DRR 18-1579 DRR 21-1229 DRR 20-0988 DRR 21-0966 DRR 21-0966 7/7/21 10/22/18 11/3/21 10/24/20 7/7/21 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 93 DRR 22-1363 2/2/23 B 3.8.1-3 B 3.8.1-4 95 96 N/A N/A 4/25/24 12/11/2024 B 3.8.1-5 93 DRR 22-1363 2/2/23 B 3.8.1-6 93 DRR 22-1363 2/2/23 B 3.8.1-7 93 DRR 22-1363 2/2/23 B 3.8.1-8 93 DRR 22-1363 2/2/23 B 3.8.1-9 93 DRR 22-1363 2/2/23 B 3.8.1-10 93 DRR 22-1363 2/2/23 B 3.8.1-11 93 DRR 22-1363 2/2/23 B 3.8.1-12 93 DRR 22-1363 2/2/23 B 3.8.1-13 93 DRR 22-1363 2/2/23 B 3.8.1-14 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23

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Wolf Creek - Unit 1 xiii Revision 96 TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)

B 3.8.1-18 95 N/A 4/25/24 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 93 DRR 22-1363 2/2/23 B 3.8.1-22 89 DRR 21-0966 7/7/21 B 3.8.1-23 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23 B 3.8.1-29 96 N/A 12/11/2024 B 3.8.1-30 96 N/A 12/11/2024 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 93 DRR 22-1363 2/2/23 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 92 DRR 22-0767 11/3/22 B 3.8.2-4 92 DRR 22-0767 11/3/22 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 90 DRR 21-1229 11/3/21 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 93 DRR 22-1363 2/2/23 B 3.8.4-4 0

Amend. No. 123 12/18/99 B 3.8.4-5 93 DRR 22-1363 2/2/23 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 93 DRR 22-1363 2/2/23 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 95 N/A 4/25/24 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

LIST OF EFFECTIVE PAGES - TECHNICAL SPECIFICATION BASES PAGE (1)

REVISION NO. (2)

CHANGE DOCUMENT (3)

DATE EFFECTIVE/

IMPLEMENTED (4)

Wolf Creek - Unit 1 xiv Revision 96 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 93 DRR 22-1363 2/2/23 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

LIST OF EFFECTIVE PAGES - TECHNICAL SPECIFICATION BASES PAGE (1)

REVISION NO. (2)

CHANGE DOCUMENT (3)

DATE EFFECTIVE/

IMPLEMENTED (4)

Wolf Creek - Unit 1 xv Revision 96 TAB - B 3.9 REFUELING OPERATIONS (continued)

B 3.9.6-2 90 DRR 21-1229 11/3/21 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. As of Revision 94, the TS Bases is no longer controlled in PMAC.

Therefore, starting with Revision 94, the DRR number will be N/A.

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

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-1 Revision 96 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature).

Therefore, with a negative MTC, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.

MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by surveillances.

Reload cores are designed so that the beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis and fuel cycles are evaluated to ensure that the MTC does not exceed the EOC limit.

The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the USAR accident and transient analyses.

If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity.

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-2 Revision 96 BASES BACKGROUND The SRs for verification of the MTC at the beginning and near the end (continued) of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.

APPLICABLE The acceptance criteria for the specified MTC are:

SAFETY ANALYSES

a.

The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2); and

b.

The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

The USAR, Chapter 15 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).

The consequences of accidents that cause core overheating must be evaluated when the MTC is positive (part-power conditions) or zero (full-power conditions). Such accidents include the rod withdrawal transients from either subcritical or at-power conditions, loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.

In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2).

MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC verification is conducted for conditions of an RCS boron concentration equivalent to 300 ppm at an equilibrium, all rods out, RTP condition. The value determined during the verification may be extrapolated to project the EOC value, in order to confirm reload design predictions.

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-3 Revision 96 BASES APPLICABLE MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). Even though it is not SAFETY ANALYSES directly observed and controlled from the control room, MTC is (continued) considered an initial condition process variable because of its dependence on boron concentration.

LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation.

Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and less negative than a given lower bound. The MTC is most positive near BOC; this upper bound must not be exceeded. This maximum upper limit occurs near BOC, all rods out (ARO), hot zero power conditions. At EOC the MTC takes on its most negative value, when the lower bound becomes important. This LCO exists to ensure that both the upper and lower bounds are not exceeded.

During operation, the LCO is ensured through surveillances. The Surveillance checks at BOC and EOC on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met.

The LCO establishes a maximum positive value that cannot be exceeded.

The BOC positive limit and the EOC negative limit are established in the COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule.

APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above.

In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled CONTROL ROD assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-4 Revision 96 BASES APPLICABILITY not applicable, since no Design Basis Accidents using the MTC as an (continued) analysis assumption are initiated from these MODES.

ACTIONS A.1 If the BOC MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits.

The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides enough time for evaluating the MTC surveillance and computing the required bank withdrawal limits.

As cycle burnup is increased, the RCS boron concentration will be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined.

At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect.

B.1 If the required administrative withdrawal limits at BOC are not established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the unit must be brought to MODE 2 with keff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses. Taking the plant to MODE 2 with Keff < 1.0 is for the purpose of meeting the LCO Mode Applicability requirements within the allowed Completion Time. In accordance with plant procedures, the plant is brought to MODE 3 to allow for more stable plant conditions prior to resumption of power operation.

The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

C.1 Exceeding the EOC MTC limit means that the safety analysis assumptions for the EOC accidents that use a bounding negative MTC value may be invalid. If the EOC MTC limit is exceeded, the plant must be brought to a MODE or condition in which the LCO requirements are

MTC B 3.1.3 Wolf Creek - Unit 1 B 3.1.3-5 Revision 96 BASES ACTIONS C.1 (continued) not applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Time 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 3.1.3.1 REQUIREMENTS This SR requires verification of the MTC at BOC prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO.

Meeting the BOC limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.

If required, the BOC MTC can be used to establish administrative withdrawal limits for control banks.

SR 3.1.3.2 In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOC full power conditions. This surveillance may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOC LCO limit. The 300 ppm SR value is sufficiently less negative than the EOC LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.

SR 3.1.3.2 is modified by three Notes that include the following requirements:

1.

The SR is required to be performed once each cycle within 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.

Containment B 3.6.1 Wolf Creek - Unit 1 B 3.6.1-4 Revision 96 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 in an outage when performing only Type B and C testing, 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

Containment B 3.6.1 Wolf Creek - Unit 1 B 3.6.1-5 Revision 96 BASES 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.

EES B 3.7.13 Wolf Creek - Unit 1 B 3.7.13-5 Revision 96 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 8.

AC Sources - Operating B 3.8.1 Wolf Creek - Unit 1 B 3.8.1-4 Revision 96 BASES LCO Switchyard transformers #10 and #11 and ESF transformers XNB01 and (continued)

XNB02 have automatic load tap changers (LTCs). For an ESF transformer to be OPERABLE, its LTC and, for XNB01, the LTC for the transformer feeding it, must be capable of meeting specified safety design functions identified in calculation XX-E-006.

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 met from a variety of initial conditions such as DG in standby with the 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 (to the extent practical) 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. A circuit may be connected to more than one ESF bus through the normal or alternate feeder breaker and not violate separation criteria. In this alignment, the offsite circuit not connected to an ESF bus is required to be considered inoperable. The applicable ACTIONS of this specification must be taken.

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

AC Sources - Operating B 3.8.1 Wolf Creek - Unit 1 B 3.8.1-29 Revision 96 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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. For the alternate power source, this testing may include any series of sequential, overlapping, or total steps so that the ability to manually synchronize and transfer loads 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, 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

AC Sources - Operating B 3.8.1 Wolf Creek - Unit 1 B 3.8.1-30 Revision 96 BASES SURVEILLANCE SR 3.8.1.16 (continued)

REQUIREMENTS 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, 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

ML25058A069

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