ML26064A240
| ML26064A240 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/05/2026 |
| From: | Hamman D Wolf Creek |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 001257 | |
| Download: ML26064A240 (0) | |
Text
P.O. Box 411 l Burlington, KS 66839 l 620-364-8831 Dustin T. Hamman Director Nuclear and Regulatory Affairs March 5, 2026 001257 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 97 and 98 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 97 and 98) under the provisions to TS Section 5.5.14 and a List of Effective Pages. This submittal reflects changes from January 1, 2025, through December 31, 2025.
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 26-001257
Enclosure to 001257 Wolf Creek Generating Station Changes to the Technical Specification Bases (64 pages)
TS BASES REVISION: 97 TECHNICAL SPECIFICATION BASES Wolf Creek Generating Station, Unit 1 Summary of Revision 97:
- 1)
Revised TS Bases pages B 3.3.1-56, B 3.3.1-57, B 3.3.2-50, and B 3.3.2-52, as part of incorporation of License Amendment 243. Amendment 243 revises technical specification definitions for engineered safety features (ESF) response time and reactor trip system (RTS) response time that are referenced in surveillance requirements (SRs). These changes are based on Technical Specifications Task Force (TSTF) traveler TSTF-569, Revision 2, Revise Response Time Testing Definition, dated June 25, 2019.
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Wolf Creek - Unit 1 i Revision 97 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)
B 2.1.1-1 0
Amend. No. 123 12/18/99 B 2.1.1-2 14 DRR 03-0102 2/12/03 B 2.1.1-3 14 DRR 03-0102 2/12/03 B 2.1.1-4 14 DRR 03-0102 2/12/03 B 2.1.2-1 84 DRR 20-0400 08/18/20 B 2.1.2-2 84 DRR 20-0400 08/18/20 B 2.1.2-3 81 DRR 19-1027 10/28/19 TAB - B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 81 DRR 19-1027 10/28/19 B 3.0-2 0
Amend. No. 123 12/18/99 B 3.0-3 81 DRR 19-1027 10/28/19 B 3.0-4 81 DRR 19-1027 10/28/19 B 3.0-5 81 DRR 19-1027 10/28/19 B 3.0-6 81 DRR 19-1027 10/28/19 B 3.0-7 81 DRR 19-1027 10/28/19 B 3.0-8 81 DRR 19-1027 10/28/19 B 3.0-9 81 DRR 19-1027 10/28/19 B 3.0-10 81 DRR 19-1027 10/28/19 B 3.0-11 81 DRR 19-1027 10/28/19 B 3.0-12 81 DRR 19-1027 10/28/19 B 3.0-13 81 DRR 19-1027 10/28/19 B 3.0-14 81 DRR 19-1027 10/28/19 B 3.0-15 81 DRR 19-1027 10/28/19 B 3.0-16 81 DRR 19-1027 10/28/19 B 3.0-17 81 DRR 19-1027 10/28/19 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1-1 0
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Wolf Creek - Unit 1 ii Revision 97 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)
B 3.1.3-4 96 N/A 12/11/2024 B 3.1.3-5 96 N/A 12/11/2024 B 3.1.3-6 95 N/A 4/25/24 B 3.1.4-1 0
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Amend. No. 123 12/18/99 B 3.1.8-3 15 DRR 03-0860 7/10/03 B 3.1.8-4 15 DRR 03-0860 7/10/03 B 3.1.8-5 89 DRR 21-0966 7/7/21 B 3.1.8-6 89 DRR 21-0966 7/7/21 B 3.1.9-1 84 DRR 20-0400 08/18/20 B 3.1.9-2 84 DRR 20-0400 08/18/20 B 3.1.9-3 84 DRR 20-0400 08/18/20 B 3.1.9-4 84 DRR 20-0400 08/18/20 B 3.1.9-5 89 DRR 21-0966 7/7/21 TAB - B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1-1 48 DRR 10-3740 12/28/10 B 3.2.1-2 0
Amend. No. 123 12/18/99 B 3.2.1-3 48 DRR 10-3740 12/28/10 B 3.2.1-4 48 DRR 10-3740 12/28/10 B 3.2.1-5 48 DRR 10-3740 12/28/10 B 3.2.1-6 48 DRR 10-3740 12/28/10 B 3.2.1-7 48 DRR 10-3740 12/28/10 B 3.2.1-8 89 DRR 21-0966 7/7/21 B 3.2.1-9 89 DRR 21-0966 7/7/21 B 3.2.1-10 70 DRR 15-0944 4/28/15
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Wolf Creek - Unit 1 iii Revision 97 TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued)
B 3.2.2-1 48 DRR 10-3740 12/28/10 B 3.2.2-2 0
Amend. No. 123 12/18/99 B 3.2.2-3 48 DRR 10-3740 12/28/10 B 3.2.2-4 48 DRR 10-3740 12/28/10 B 3.2.2-5 48 DRR 10-3740 12/28/10 B 3.2.2-6 89 DRR 21-0966 7/7/21 B 3.2.3-1 0
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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 97 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 97 N/A 5/27/25 B 3.3.1-57 97 N/A 5/27/25 B 3.3.1-58 97 N/A 5/27/25 B 3.3.1-59 97 N/A 5/27/25 B 3.3.1-60 97 N/A 5/27/25 B 3.3.1-61 B 3.3.1-62 (new) 97 97 N/A N/A 5/27/25 5/27/25 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 97 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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Wolf Creek - Unit 1 vi Revision 97 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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Wolf Creek - Unit 1 vii Revision 97 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)
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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
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Amend. No. 123 12/18/99
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Wolf Creek - Unit 1 viii Revision 97 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
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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
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Wolf Creek - Unit 1 ix Revision 97 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
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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
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Wolf Creek - Unit 1 x Revision 97 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
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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 97 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
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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 97 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 97 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
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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
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Wolf Creek - Unit 1 xiv Revision 97 TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)
B 3.8.6-5 0
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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
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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
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IMPLEMENTED (4)
Wolf Creek - Unit 1 xv Revision 97 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.
RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-56 Revision 97 BASES SURVEILLANCE SR 3.3.1.16 (continued)
REQUIREMENTS channels with response times listed as N.A. No response time testing requirements apply where N.A. is listed in Table B 3.3.1-2. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor until loss of stationary gripper coil voltage.
For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time verification is performed with the time constants set at their nominal values. The response time may be measured by a series of overlapping tests, or other verification (e.g.,
Ref. 7), such that the entire response time is measured.
Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated response times with actual response time tests on the remainder of the channel. Allocations for response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements (Ref. 7), provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, (Ref. 14) provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 6 and Ref. 14, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection
RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-57 Revision 97 BASES SURVEILLANCE SR 3.3.1.16 (continued)
REQUIREMENTS Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 15).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal.
Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input to the first electronic component in the channel.
REFERENCES
- 1.
USAR, Chapter 7.
- 2.
USAR, Chapter 15.
- 3.
- 4.
10 CFR 50.49.
- 5.
WCNOC Nuclear Safety Analysis Setpoint Methodology for the Reactor Protection System, (TR-89-0001).
- 6.
WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996.
- 7.
WCAP-9226, Reactor Core Response to Excessive Secondary Steam Releases, Revision 1, January 1978.
- 8.
IE Information Notice 79-22, Qualification of Control Systems, September 14, 1979.
- 9.
Wolf Creek Setpoint Methodology Report, SNP(KG)-492, August 29, 1984.
- 10.
USAR, Table 15.0-4.
- 11.
WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, October 1998.
RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-58 Revision 97 BASES REFERENCES
- 12.
WCAP-15376-P-A, Revision 1, Risk-Informed Assessment of the (continued)
RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times, March 2003.
- 13.
WOG-06-17, WCAP-10271-P-A Justification for Bypass Test Time and Completion Time Technical Specification Changes for Reactor Trip on Turbine Trip (ITSWG Action item #314), January 20, 2006.
- 14.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, December 1995.
- 15. to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing.
RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-59 Revision 97 TABLE B 3.3.1-1 (Page 1 of 2)
FUNCTION TRIP SETPOINT(a)
- 1.
- 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 97 TABLE B 3.3.1-1 (Page 2 of 2)
FUNCTION TRIP SETPOINT(a)
- 18. Reactor Trip System Interlocks
- a. Intermediate Range Neutron Flux, P-6
- b. Low Power Reactor Trips Block, P-7
- c. Power Range Neutron Flux, P-8
- d. Power Range Neutron Flux, P-9
- e. Power Range Neutron Flux, P-10
- f.
Turbine Impulse Pressure, P-13
- 19. Reactor Trip Breakers
- 20. Reactor Trip breaker Undervoltage and Shunt Trip Mechanisms
- 21. Automatic Trip Logic N.A.
1.0E-10 amps N.A.
48% RTP 50% RTP 10% RTP 10% turbine power N.A.
N.A.
N.A.
(a) The inequality sign only indicates conservative direction. The as-left value will be within a two-sided calibration tolerance band on either side of the nominal value. This also applies to the Overtemperature ¨T and Overpower ¨T K and values.
RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-61 Revision 97 TABLE B 3.3.1-2 (Page 1 of 2)
FUNCTIONAL UNIT RESPONSE TIME
- 1.
- 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.
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.
RTS Instrumentation B 3.3.1 Wolf Creek - Unit 1 B 3.3.1-62 Revision 97 TABLE B 3.3.1-2 (Page 2 of 2)
FUNCTIONAL UNIT RESPONSE TIME
- 16. Turbine Trip
- a. Low Fluid Oil Pressure
- b. Turbine Stop Valve Closure
- 17. Safety Injection Input for ESF
- 18. Reactor Trip System Interlocks
- 19. Reactor Trip Breakers
- 20. Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms
- 21. Automatic Trip and Interlock Logic N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
ESFAS Instrumentation B 3.3.2 Wolf Creek - Unit 1 B 3.3.2-50 Revision 97 BASES SURVEILLANCE SR 3.3.2.10 (continued)
REQUIREMENTS response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, (Ref. 15) provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 8 and Ref. 15, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 16).
The NRC approved the use of ASME Code Case OMN-1, Alternative Rules for Preservice and Inservice Testing of Certain Electric Motor-Operated Valve Assemblies in Light-Water Reactor Plants, as an alternative to stroke time testing for motor-operated valves (Ref. 14). The parameters that must be present to achieve the analyzed response time under design basis conditions are measured to ensure the valve is capable of performing its safety function. This process verifies design basis capability, including response time, and is a significant improvement over simple stroke time measurement. This process allows the establishment of periodic valve test intervals if there is assurance that the valve will remain capable of performing its safety function throughout the interval.
ESF response times specified in Table B 3.3.2-2 which include sequential operation of RWST and VCT valves (Notes 3 and 4) are based on values assumed in the non-LOCA safety analyses. These analyses take credit for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves.
ESFAS Instrumentation B 3.3.2 Wolf Creek - Unit 1 B 3.3.2-51 Revision 97 BASES SURVEILLANCE SR 3.3.2.10 (continued)
REQUIREMENTS When the sequential operation of the RWST and VCT valves is not included in the response times (Note 7), the values specified are based on the LOCA analyses. The LOCA analyses take credit for injection flow regardless of the source. Verification of the response times specified in Table B 3.3.2-2 will assure that the assumptions used for the LOCA and non-LOCA analyses with respect to the operation of the VCT and RWST valves are valid.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 900 psig in the SGs.
SR 3.3.2.11 3.3.2.11 is the performance of a TADOT as described in SR 3.3.2.8, except that it is performed for the P-4 Reactor Trip Interlock. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint. This TADOT does not include the circuitry associated with steam dump operation since it is control grade circuitry.
SR 3.3.2.12 SR 3.3.2.12 is the performance of a monthly COT on ESFAS Function 6.h, Auxiliary Feedwater Pump Suction Transfer on Suction Pressure - Low.
A COT is performed to ensure the channel will perform the intended Function. Setpoints must be found within the Allowable Values specified in Table 3.3.2-1.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
ESFAS Instrumentation B 3.3.2 Wolf Creek - Unit 1 B 3.3.2-52 Revision 97 BASES REFERENCES
- 1.
USAR, Chapter 6.
- 2.
USAR, Chapter 7.
- 3.
USAR, Chapter 15.
- 4.
- 5.
10 CFR 50.49.
- 6.
WCNOC Nuclear Safety Analysis Setpoint Methodology for the Reactor Protection System, TR-89-0001.
- 7.
WCAP-10271-P-A Supplement 2, Revision 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," June 1990.
- 8.
WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996.
- 9.
Wolf Creek Setpoint Methodology Report, SNP (KG)-492, August 29, 1984.
- 10.
Amendment No. 43 to Facility Operating License No. NPF-42, March 29, 1991.
- 11.
WCAP-14333-P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, October 1998.
- 12.
10 CFR 50.55a(b)(3)(iii), Code Case OMN-1.
- 13.
Performance Improvement Request (PIR) 2005-2067.
- 14.
Amendment No. 231 to Renewed Facility Operating License No.
NPF-42, February 23, 2022.
- 15.
WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, December 1995.
- 16. to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)
Response Time Testing.
TS BASES REVISION: 98 TECHNICAL SPECIFICATION BASES Wolf Creek Generating Station, Unit 1 Summary of Revision 98:
- 1)
Revised TS Bases Section B 3.2.1 as part of incorporation of License Amendment 244.
Amendment 244 revises TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology), to implement the methodology in Pressurized Water Reactor Owners Group Topical Report WCAP-17661-P-A, Revision 1, Improved RAOC [Relaxed Axial Offset Control] and CAOC
[Constant Axial Offset Control] FQ Surveillance Technical Specifications.
- 2)
Revised TS Bases B 3.4.15 to remove the requirement that the plant process computer must be capable of calculating the leakage rate indicated by increases in sump level. This will allow for manually calculating RCS leak rate once per shift when the plant process computer is out of service.
This is consistent with the Wolf Creek Technical Specifications (TS) prior to the adoption of the Improved TS (ITS), when Surveillance Requirement 4.4.6.2.1.b was deleted. Because this was an approved method for meeting the LCO (current ITS LCO is 3.4.15) prior to adoption of ITS, this method may still be used to meet the LCO,
- 3)
Revised TS Bases page B 3.8.1-1 to indicate that with the addition of the Blackberry line, offsite power is now supplied to the unit switchyard from the transmission network by four transmission lines.
LIST OF EFFECTIVE PAGES - TECHNICAL SPECIFICATION BASES PAGE (1)
REVISION NO. (2)
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DATE EFFECTIVE/
IMPLEMENTED (4)
Wolf Creek - Unit 1 i Revision 98 TAB - Title Page Technical Specification Cover Page Title Page TAB - Table of Contents i
98 N/A 10/15/25 ii 81 DRR 19-1027 10/28/19 iii 81 DRR 19-1027 10/28/19 TAB - B 2.0 SAFETY LIMITS (SLs)
B 2.1.1-1 0
Amend. No. 123 12/18/99 B 2.1.1-2 14 DRR 03-0102 2/12/03 B 2.1.1-3 14 DRR 03-0102 2/12/03 B 2.1.1-4 14 DRR 03-0102 2/12/03 B 2.1.2-1 84 DRR 20-0400 08/18/20 B 2.1.2-2 84 DRR 20-0400 08/18/20 B 2.1.2-3 81 DRR 19-1027 10/28/19 TAB - B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILTY B 3.0-1 81 DRR 19-1027 10/28/19 B 3.0-2 0
Amend. No. 123 12/18/99 B 3.0-3 81 DRR 19-1027 10/28/19 B 3.0-4 81 DRR 19-1027 10/28/19 B 3.0-5 81 DRR 19-1027 10/28/19 B 3.0-6 81 DRR 19-1027 10/28/19 B 3.0-7 81 DRR 19-1027 10/28/19 B 3.0-8 81 DRR 19-1027 10/28/19 B 3.0-9 81 DRR 19-1027 10/28/19 B 3.0-10 81 DRR 19-1027 10/28/19 B 3.0-11 81 DRR 19-1027 10/28/19 B 3.0-12 81 DRR 19-1027 10/28/19 B 3.0-13 81 DRR 19-1027 10/28/19 B 3.0-14 81 DRR 19-1027 10/28/19 B 3.0-15 81 DRR 19-1027 10/28/19 B 3.0-16 81 DRR 19-1027 10/28/19 B 3.0-17 81 DRR 19-1027 10/28/19 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1-1 0
Amend. No. 123 12/18/99 B 3.1.1-2 0
Amend. No. 123 12/18/99 B 3.1.1-3 95 N/A 4/25/24 B 3.1.1-4 81 DRR 19-1027 10/28/19 B 3.1.1-5 89 DRR 21-0966 7/7/21 B 3.1.2-1 0
Amend. No. 123 12/18/99 B 3.1.2-2 0
Amend. No. 123 12/18/99 B 3.1.2-3 0
Amend. No. 123 12/18/99 B 3.1.2-4 0
Amend. No. 123 12/18/99 B 3.1.2-5 89 DRR 21-0966 7/7/21 B 3.1.3-1 96 N/A 12/11/2024 B 3.1.3-2 96 N/A 12/11/2024 B 3.1.3-3 96 N/A 12/11/2024
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Wolf Creek - Unit 1 ii Revision 98 TAB - B 3.1 REACTIVITY CONTROL SYSTEMS (continued)
B 3.1.3-4 96 N/A 12/11/2024 B 3.1.3-5 96 N/A 12/11/2024 B 3.1.3-6 95 N/A 4/25/24 B 3.1.4-1 0
Amend. No. 123 12/18/99 B 3.1.4-2 0
Amend. No. 123 12/18/99 B 3.1.4-3 48 DRR 10-3740 12/28/10 B 3.1.4-4 0
Amend. No. 123 12/18/99 B 3.1.4-5 0
Amend. No. 123 12/18/99 B 3.1.4-6 48 DRR 10-3740 12/28/10 B 3.1.4-7 0
Amend. No. 123 12/18/99 B 3.1.4-8 89 DRR 21-0966 7/7/21 B 3.1.4-9 89 DRR 21-0966 7/7/21 B 3.1.5-1 0
Amend. No. 123 12/18/99 B 3.1.5-2 0
Amend. No. 123 12/18/99 B 3.1.5-3 0
Amend. No. 123 12/18/99 B 3.1.5-4 89 DRR 21-0966 7/7/21 B 3.1.6-1 0
Amend. No. 123 12/18/99 B 3.1.6-2 0
Amend. No. 123 12/18/99 B 3.1.6-3 0
Amend. No. 123 12/18/99 B 3.1.6-4 0
Amend. No. 123 12/18/99 B 3.1.6-5 89 DRR 21-0966 7/7/21 B 3.1.6-6 95 N/A 4/25/24 B 3.1.7-1 0
Amend. No. 123 12/18/99 B 3.1.7-2 0
Amend. No. 123 12/18/99 B 3.1.7-3 48 DRR 10-3740 12/28/10 B 3.1.7-4 48 DRR 10-3740 12/28/10 B 3.1.7-5 48 DRR 10-3740 12/28/10 B 3.1.7-6 0
Amend. No. 123 12/18/99 B 3.1.8-1 0
Amend. No. 123 12/18/99 B 3.1.8-2 0
Amend. No. 123 12/18/99 B 3.1.8-3 15 DRR 03-0860 7/10/03 B 3.1.8-4 15 DRR 03-0860 7/10/03 B 3.1.8-5 89 DRR 21-0966 7/7/21 B 3.1.8-6 89 DRR 21-0966 7/7/21 B 3.1.9-1 84 DRR 20-0400 08/18/20 B 3.1.9-2 84 DRR 20-0400 08/18/20 B 3.1.9-3 84 DRR 20-0400 08/18/20 B 3.1.9-4 84 DRR 20-0400 08/18/20 B 3.1.9-5 89 DRR 21-0966 7/7/21 TAB - B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1-1 98 N/A 10/15/25 B 3.2.1-2 98 N/A 10/15/25 B 3.2.1-3 98 N/A 10/15/25 B 3.2.1-4 98 N/A 10/15/25 B 3.2.1-5 98 N/A 10/15/25 B 3.2.1-6 98 N/A 10/15/25 B 3.2.1-7 98 N/A 10/15/25 B 3.2.1-8 98 N/A 10/15/25 B 3.2.1-9 98 N/A 10/15/25 B 3.2.1-10 98 N/A 10/15/25
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Wolf Creek - Unit 1 iii Revision 98 TAB - B 3.2 POWER DISTRIBUTION LIMITS (continued)
B 3.2.1-11 (new) 98 N/A 10/15/25 B 3.2.1-12 (new) 98 N/A 10/15/25 B 3.2.1-13 (new) 98 N/A 10/15/25 B 3.2.1-14 (new) 98 N/A 10/15/25 B 3.2.2-1 48 DRR 10-3740 12/28/10 B 3.2.2-2 0
Amend. No. 123 12/18/99 B 3.2.2-3 48 DRR 10-3740 12/28/10 B 3.2.2-4 48 DRR 10-3740 12/28/10 B 3.2.2-5 48 DRR 10-3740 12/28/10 B 3.2.2-6 89 DRR 21-0966 7/7/21 B 3.2.3-1 0
Amend. No. 123 12/18/99 B 3.2.3-2 0
Amend. No. 123 12/18/99 B 3.2.3-3 89 DRR 21-0966 7/7/21 B 3.2.4-1 0
Amend. No. 123 12/18/99 B 3.2.4-2 0
Amend. No. 123 12/18/99 B 3.2.4-3 48 DRR 10-3740 12/28/10 B 3.2.4-4 0
Amend. No. 123 12/18/99 B 3.2.4-5 48 DRR 10-3740 12/28/10 B 3.2.4-6 89 DRR 21-0966 7/7/21 B 3.2.4-7 89 DRR 21-0966 7/7/21 TAB - B 3.3 INSTRUMENTATION B 3.3.1-1 84 DRR 20-0400 08/18/20 B 3.3.1-2 0
Amend. No. 123 12/18/99 B 3.3.1-3 0
Amend. No. 123 12/18/99 B 3.3.1-4 0
Amend. No. 123 12/18/99 B 3.3.1-5 0
Amend. No. 123 12/18/99 B 3.3.1-6 0
Amend. No. 123 12/18/99 B 3.3.1-7 5
DRR 00-1427 10/12/00 B 3.3.1-8 0
Amend. No. 123 12/18/99 B 3.3.1-9 84 DRR 20-0400 08/18/20 B 3.3.1-10 84 DRR 20-0400 08/18/20 B 3.3.1-11 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
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Wolf Creek - Unit 1 iv Revision 98 TAB - B 3.3 INSTRUMENTATION (continued)
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Wolf Creek - Unit 1 vi Revision 98 TAB - B 3.3 INSTRUMENTATION (continued)
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Wolf Creek - Unit 1 viii Revision 98 TAB - B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)
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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 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
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Wolf Creek - Unit 1 xii Revision 98 TAB - B 3.7 PLANT SYSTEMS (continued)
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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
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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 98 N/A 10/15/25 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
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Wolf Creek - Unit 1 xiii Revision 98 TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)
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 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
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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
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Wolf Creek - Unit 1 xiv Revision 98 TAB - B 3.8 ELECTRICAL POWER SYSTEMS (continued)
B 3.8.6-1 0
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REVISION NO. (2)
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Wolf Creek - Unit 1 xv Revision 98 TAB - B 3.9 REFUELING OPERATIONS (continued)
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 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.
Wolf Creek - Unit 1 i
Revision 98 TABLE OF CONTENTS B 2.0 SAFETY LIMITS (SLs)................................................................................... B 2.1.1-1 B 2.1.1 Reactor Core SLs........................................................................... B 2.1.1-1 B 2.1.2 Reactor Coolant System (RCS) Pressure SL................................. B 2.1.2-1 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY................. B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.............................. B 3.0-12 B 3.1 REACTIVITY CONTROL SYSTEMS...................................................... B 3.1.1-1 B 3.1.1 SHUTDOWN MARGIN (SDM)...................................................... B 3.1.1-1 B 3.1.2 Core Reactivity............................................................................... B 3.1.2-1 B 3.1.3 Moderator Temperature Coefficient (MTC)..................................... B 3.1.3-1 B 3.1.4 Rod Group Alignment Limits........................................................... B 3.1.4-1 B 3.1.5 Shutdown Bank Insertion Limits..................................................... B 3.1.5-1 B 3.1.6 Control Bank Insertion Limits.......................................................... B 3.1.6-1 B 3.1.7 Rod Position Indication................................................................... B 3.1.7-1 B 3.1.8 PHYSICS TESTS Exceptions - MODE 2........................................ B 3.1.8-1 B 3.1.9 RCS Boron Limitations < 500°F B 3.1.9-1 B 3.2 POWER DISTRIBUTION LIMITS........................................................... B 3.2.1-1 B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))
(RAOC - T(Z) Methodology)..................................................... B 3.2.1-1 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH )............................................................................. B 3.2.2-1 B 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology)....................................... B 3.2.3-1 B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)................................... B 3.2.4-1 B 3.3 INSTRUMENTATION............................................................................. B 3.3.1-1 B 3.3.1 Reactor Trip System (RTS) Instrumentation................................... B 3.3.1-1 B 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation............................................. B 3.3.2-1 B 3.3.3 Post Accident Monitoring (PAM) Instrumentation........................... B 3.3.3-1 B 3.3.4 Remote Shutdown System............................................................. B 3.3.4-1 B 3.3.5 Loss of Power (LOP) Diesel Generator (DG)
Start Instrumentation................................................................ B 3.3.5-1 B 3.3.6 Containment Purge Isolation Instrumentation......................................................................... B 3.3.6-1
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-1 Revision 98 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) (RAOC - T(Z) Methodology)
BASES BACKGROUND The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core.
FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ(Z) is a measure of the peak fuel pellet power within the reactor core.
During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.4, Rod Group Alignment Limits, LCO 3.1.5, Shutdown Bank Insertion Limits, and LCO 3.1.6, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis.
FQ(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution.
FQ(Z) is not directly measurable but is inferred from a power distribution measurement obtained with either the movable incore detector system or the Power Distribution Monitoring System (PDMS). The results of the three-dimensional power distribution measurement are analyzed to derive a measured value for FQ(Z). These measurements are generally taken with the core at or near equilibrium conditions. However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ(Z) that are present during nonequilibrium situations, such as load following.
To account for these possible variations, the elevation dependent measured planar radial peaking factors, FXY(Z), are increased by an elevation dependent factor, [T(Z)]COLR, that accounts for the expected maximum values of the transient axial power shapes postulated to occur during RAOC operation. Thus, [T(Z)]COLR accounts for the worst case non-equilibrium power shapes that are expected for the assumed RAOC operating space.
The RAOC operating space is defined as the combination of AFD and Control Bank Insertion Limits assumed in the calculation of a particular
[T(Z)]COLR function. The [T(Z)]COLR factors are directly dependent on the
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-2 Revision 98 BASES BACKGROUND AFD and Control Bank Insertion Limit assumptions. The COLR may (continued) contain different [T(Z)]COLR functions that reflect different operating space assumptions. If the limit on FQ(Z) is exceeded, a more restrictive operating space may be implemented to gain margin for future non-equilibrium operation.
Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion.
APPLICABLE This LCO precludes core power distributions that violate the following fuel SAFETY ANALYSES design criteria:
- a.
During a large break loss of coolant accident (LOCA), the peak cladding temperature must not exceed 2200°F (Ref. 1);
- b.
During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition;
- c.
During an ejected rod accident, the average fuel pellet enthalpy at the hot spot in irradiated fuel must not exceed 200 cal/gm (Ref. 2);
and
- d.
The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
Limits on FQ(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the LOCA peak cladding temperature is typically most limiting.
FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ(Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents.
FQ(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-3 Revision 98 BASES LCO The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships:
(
)
(
)
(
)
(
)
0.5 P
for Z
K 0.5 CFQ Z
F 0.5 P
for Z
K P
CFQ Z
F Q
Q
where:
CFQ = FQRTP is the FQ(Z) limit at RTP provided in the COLR, K(Z) is the normalized FQ(Z) limit as a function of core height provided in the COLR, and RTP POWER THERMAL
=
P The actual values of CFQ and K(Z) are given in the COLR.
For Relaxed Axial Offset Control operation, FQ(Z) is approximated by FQC(Z) and FQW(Z). Thus, both FQC(Z) and FQW(Z) must meet the preceding limits on FQ(Z).
An FQC(Z) evaluation requires obtaining a power distribution measurement in MODE 1, from which we obtain the measured value (FQM(Z)) of FQ(Z).
If the power distribution measurement is obtained with the movable incore detector system, FQC(Z) = FQM(Z) (1.03) (1.05) = FQM(Z) (1.0815) where 1.03 is a factor that accounts for fuel manufacturing tolerances and 1.05 is a factor that accounts for flux map measurement uncertainty.
(Ref. 4)
If the power distribution measurement is obtained with the Power Distribution Monitoring System, FQC(Z) = FQM(Z) (1.03) (1.00 + UQ/100) where 1.03 is a factor that accounts for fuel manufacturing tolerances and UQ is a factor that accounts for Power Distribution Monitoring System measurement uncertainty (%), determined as described in Reference 6.
FQC(Z) is an excellent approximation for FQ(Z) when the reactor is at the steady state power at which the power distribution measurement was taken.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-4 Revision 98 BASES LCO The expression for FQW(Z) is:
(continued)
FQ WZ=FXY MZTZ P
COLR AXYZRj1.0815 for P>0.5 FQ WZ=FXY MZTZ 0.5 COLR AXYZRj1.0815 for P0.5 The various factors in this expression are defined below:
FXYM(Z) is the measured radial peaking factor at axial location Z and is equal to the value of FQM(Z)/PM(Z), where PM(Z) is the measured core average axial power shape.
[T(Z)]COLR is the cycle and burnup dependent function, specified in the COLR, which accounts for power distribution transients encountered during non-equilibrium normal operation. [T(Z)]COLR functions are specified for each analyzed RAOC operating space (i.e., each unique combination of AFD limits and Control Bank Insertion Limits). The
[T(Z)]COLR functions account for the limiting non-equilibrium axial power shapes postulated to occur during normal operation for each RAOC operating space. Limiting power shapes at both full and reduced power operation are considered in determining the maximum values of
[T(Z)]COLR. The [T(Z)]COLR functions also account for the following effects:
(1) the presence of spacer grids in the fuel assembly, (2) the increase in radial peaking in rodded core planes due to the presence of control rods during non-equilibrium normal operation, (3) the increase in radial peaking that occurs during part-power operation due to reduced fuel and moderator temperatures, and (4) the increase in radial peaking due to non-equilibrium xenon effects. The [T(Z)]COLR functions are normally calculated assuming that the Surveillance is performed at nominal RTP conditions with all shutdown and control rods full withdrawn, i.e., all rods out (ARO) Surveillance specific [T(Z)]COLR values may be generated for a given surveillance core condition.
P is the THERMAL POWER / RTP.
AXY(Z) is a function that adjusts the FQW(Z) Surveillance for differences between the reference core condition assumed in generating the
[T(Z)]COLR function and the actual core condition that exists when the Surveillance is performed. Normally, this reference core condition is 100% RTP, all rods out, and equilibrium xenon. For simplicity, AXY(Z) may be assumed to be 1.0, as this will typically result in an accurate FQW(Z) Surveillance result for a Surveillance that is performed at or near the reference core condition, and an underestimation of the available margin to the FQ limit for Surveillances that are performed at core conditions different from the reference condition. Alternately, the AXY(Z)
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-5 Revision 98 BASES LCO function may be calculated using the NRC approved methodology in (continued)
Reference 7.
1.0815 is a factor that accounts for fuel manufacturing tolerances and measurement uncertainty.
Rj is a cycle and burnup dependent analytical factor specified in the COLR that accounts for potential increases in FQW(Z) between Surveillances. Rj values are provided for each RAOC operating space.
The FQ(Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during either a large or small break LOCA.
This LCO requires operation within the bounds assumed in the safety analyses. Violating the LCO limits for FQ(Z) could result in unacceptable consequences if a design basis event were to occur while FQ(Z) exceeds its specified limits. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ(Z) limits. If FQ(Z) cannot be maintained within the LCO limits, a more restrictive RAOC operating space must be implemented or core power limits and AFD limits must be reduced.
APPLICABILITY The FQ(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses.
Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.
ACTIONS A.1 Reducing THERMAL POWER by 1% RTP for each 1% by which FQC(Z) exceeds its limit, maintains an acceptable absolute power density. FQC(Z) is FQM(Z) multiplied by factors which account for manufacturing tolerances and measurement uncertainties. FQM(Z) is the measured value of FQ(Z).
The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time. The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of FQC(Z) and would require power reductions within 15 minutes of the FQC(Z) determination, if necessary to comply with the decreased maximum allowable power level.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-6 Revision 98 BASES ACTIONS A.1 (continued)
Decreases in FQC(Z) would allow increasing the maximum allowable power level and increasing power up to this revised limit.
If an FQ surveillance is performed at 100% RTP conditions, and both FQC(Z) and FQW(Z) exceed their limits, the option to reduce the THERMAL POWER limit in accordance with Required Action B.2.1 instead of implementing a new operating space in accordance with Required Action B.1.1, will result in a further power reduction after Required Action A.1 has been completed. However, this further power reduction would be permitted to occur over the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In the event the evaluated THERMAL POWER reduction in the COLR for Required Action B.2.1 did not result in a further power reduction (for example, if both Condition A and Condition B were entered at less than 100% RTP conditions), then the THERMAL POWER level established as a result of completing Required Action A.1 will take precedence, and will establish the effective operating power level limit for the unit until both Conditions A and B are exited.
A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by 1% for each 1% that THERMAL POWER is limited below RTP by Required Action A.1, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions.
The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of FQC(Z) and would require Power Range Neutron Flux - High trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of FQC(Z) determination, if necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints.
A.3 Reduction in the Overpower ¨T trip setpoints by 1% for each 1% that THERMAL POWER is limited below RTP by Required Action A.1, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Overpower ¨T trip setpoints initially determined by Required Action A.3 may be affected by subsequent determinations of FQC(Z) and
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-7 Revision 98 BASES ACTIONS A. 3 (continued) would require Overpower ¨T trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the FQC(Z) determination, if necessary to comply with the decreased maximum allowable Overpower ¨T trip setpoints. Decreases in FQC(Z) would allow increasing the maximum Overpower ¨T trip setpoints.
A.4 Verification that FQC(Z) has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions. Inherent in this action is identification of the cause of the out of limit condition and the correction of the cause to the extent necessary to allow safe operation at the higher power level.
Condition A is modified by a Note that requires Required Action A.4 to be performed whenever the Condition is entered prior to increasing THERMAL POWER above the limit of Required Action A.1. The Note also states that SR 3.2.1.2 is not required to be performed if this Condition is entered prior to THERMAL POWER exceeding 75% RTP after a refueling. This ensures that SR 3.2.1.1 and SR 3.2.1.2 (if required) will be performed prior to increasing THERMAL POWER above the limit of Required Action A.1, even when Condition A is exited prior to performing Required Action A.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to assure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.
B.1.1 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, FQW(Z), exceeds its specified limits, there exists a potential for FQC(Z) to become excessively high if a normal operational transient occurs. Implementing a more restrictive RAOC operating space, as specified in the COLR, within the allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> will restrict the AFD such that peaking factor limits will not be exceeded during non-equilibrium normal operation. Several RAOC operating spaces, representing successively smaller AFD envelopes and, optionally, shallower Control Bank Insertion Limits, may be specified in the COLR. The corresponding T(Z) functions for these operating spaces can be used to determine which RAOC operating space will result in acceptable non-equilibrium operation within the FQW(Z) limit.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-8 Revision 98 BASES ACTIONS B.1.2 (continued)
If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, FQW(Z), exceeds its specified limits, there exists a potential for FQC(Z) to become excessively high if a normal operational transient occurs. As discussed above, Required Action B.1.1 requires that a new RAOC operating space be implemented to restore FQW(Z) to within its limits. Required Action B.1.2 requires that SR 3.2.1.1 and SR 3.2.1.2 be performed if control rod motion occurs as a result of implementing the new RAOC operating space in accordance with Required Action B.1.1. The performance of SR 3.2.1.1 and SR 3.2.1.2 is necessary to assure FQ(Z) is properly evaluated after any rod motion resulting from the implementation of a new RAOC operating space in accordance with Required Action B.1.1.
B.2.1 When FQW(Z) exceeds it limit, Required Action B.2.1 may be implemented instead of Required Action B.1. Required Action B.2.1 limits THERMAL POWER to less than RTP by the amount specified in the COLR. It also requires reductions in the AFD limits by the amount specified in the COLR. This maintains an acceptable absolute power density relative to the maximum power density value assumed in the safety analyses.
If the required FQW(Z) margin improvement exceeds the margin improvement available from the pre-analyzed THERMAL POWER and AFD reductions provided in the COLR, then THERMAL POWER must be further reduced to less than or equal to 50% RTP. In this case, reducing THERMAL POWER to less than or equal to 50% RTP will provide additional margin in the transient FQ by the required change in THERMAL POWER and the increase in the FQ limit. This will ensure that the FQ limit is met during transient operation that may occur at or below 50% RTP.
The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable time to reduce the THERMAL POWER and AFD limits in an orderly manner to preclude entering an unacceptable condition during future non-equilibrium operation. The limit on THERMAL POWER initially determined by Required Action B.2.1 may be affected by subsequent determinations of FQW(Z) that are not within limit and could require power reductions within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the of the subsequent FQW(Z) determination, if necessary, to comply with the decreased THERMAL POWER limit. In short, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time for Required Action B.2.1 applies after each FQW(Z) determination. Decreases in subsequent FQW(Z) measurements while in Condition B would allow increasing the THERMAL POWER limit and increasing THERMAL POWER up to this revised limit.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-9 Revision 98 BASES ACTIONS B.2.1 (continued)
Required Action B.2.1 is modified by a Note that states Required Action B.2.4 shall be completed whenever Required Action B.2.1 is performed prior to increasing THERMAL POWER above the limit of Required Action B.2.1. Required Action B.2.4 requires the performance of SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit established by Required Action B.2.1. The Note ensures that the SRs will be performed even if Condition B may be exited prior to performing Required Action B.2.4. The performance of SR 3.2.1.1 and SR 3.2.1.2 is necessary to ensure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.
If an FQ surveillance is performed at 100% RTP conditions, and both FQC(Z) and FQW(Z) exceed their limits, the option to reduce the THERMAL POWER limit in accordance with Required Action B.2.1 instead of implementing a new operating space in accordance with Required Action B.1, will result in a further power reduction after Required Action A.1 has been completed. However, this further power reduction would be permitted to occur over the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. In the event the evaluated THERMAL POWER reduction in the COLR for Required Action B.2.1 did not result in a further power reduction (for example, if both Condition A and Condition B were entered at less than 100% RTP conditions), then the THERMAL POWER level established as a result of completing Required Action A.1 will take precedence, and will establish the effective operating power level limit for the unit until both Conditions A and B are exited.
B.2.2 A reduction of the Power Range Neutron Flux - High trip setpoints by 1% for each 1% by which the maximum allowable power is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in the THERMAL POWER limit and AFD limits in accordance with Required Action B.2.1.
The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action B.2.2 may be affected by subsequent determinations of FQW(Z) that are not within limit and could require Power Range Neutron Flux - High trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the subsequent FQW(Z) determination, if necessary, to comply with the decreased maximum allowable Power Range Neutron Flux -
High trip setpoints. In short, the 72-hour Completion Time for Required Action B.2.2 applies after each FQW(Z) determination.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-10 Revision 98 BASES ACTIONS B.2.3 (continued)
Reduction in the Overpower T trip setpoints value of K4 by 1% for each 1% by which the maximum allowable power is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in the THERMAL POWER limit and AFD limits in accordance with the Required Action B.2.1. The maximum allowable Overpower T trip setpoints initially determined by Required Action B.2.3 may be affected by subsequent determinations of FQW(Z) that are not within limit and could require Overpower T trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the subsequent FQW(Z) determination, if necessary, to comply with the decreased maximum allowable Overpower T trip setpoints. In short, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time for Required Action B.2.3 applies after each FQW(Z) determination. Decreases in subsequent FQW(Z) measurements while in Condition B would allow increasing the maximum allowable Overpower T trip setpoints.
B.2.4 Verification that FQW(Z) has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the maximum allowable power limit imposed by Required Action B.2.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.
C.1 If Required Actions A.1 through A.4 or B.1.1 through B.2.4 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable.
This is done by placing the plant in at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-11 Revision 98 BASES SURVEILLANCE SR 3.2.1.1 REQUIREMENTS Verification that FQC(Z) is within its specified limits involves increasing FQM(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FQC(Z), as described in the preceeding LCO section.
The limit with which FQC(Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.
Performing this Surveillance in MODE 1 prior to exceeding 75% RTP following a refueling ensures that some determination of FQC(Z) is made prior to achieving a significant power level where the peak linear heat rate could approach the limits assumed in the safety analyses.
If THERMAL POWER has been increased by 10% RTP since the initial or most recent determination of FQC(Z), another evaluation of this factor is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions at this higher power level (to ensure that FQC(Z) values are being reduced sufficiently with power increase to stay within the LCO limits).
Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions required to perform the Surveillance.
The allowance of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions at the increased THERMAL POWER level to complete the next FQC(Z) surveillance applies to situations where the FQC(Z) has already been measured at least once at a reduced THERMAL POWER level. The observed margin in the previous Surveillance will provide assurance that increasing power up to the next plateau will not exceed the FQ limit, and that the core is behaving as designed.
This Frequency condition is not intended to require verification of these parameters after every 10% increase in RTP above the THERMAL POWER at which the last verification was performed. It only requires verification after a THERMAL POWER is achieved for extended operation that is 10% higher than the THERMAL POWER at which FQC(Z) was last measured.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-12 Revision 98 BASES SURVEILLANCE SR 3.2.1.2 REQUIREMENTS (continued)
The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because power distribution measurements are taken at or near equilibrium conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the measurements. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation.
The measured FQ(Z) can be determined through a synthesis of the measured planar radial peaking factors, FXYM(Z), and the measured core average axial power shape, PM(Z). Thus, FQC(Z) is given by the following expression:
FQC(Z) = FXYM(Z) PM(Z)[1.0815] = FQM (Z)[1.0815]
For RAOC operation, the analytical [T(Z)]COLR functions, specified in the COLR for each RAOC operating space, are used together with the measured FXY(Z) values to estimate FQ(Z) for non-equilibrium operation within the RAOC operating space. When the FXY(Z) values are measured at HFP ARO conditions (AXY(Z) equals 1.0), FQW(Z) is given by the following expression:
FQW(Z) = FXYM(Z) [T(Z)]COLR Rj [1.0815]
Non-equilibrium operation can result in significant changes to the axial power shape. To a lesser extent, non-equilibrium operation can increase the radial peaking factors, FXY(Z), through control rod insertion and through reduced Doppler and moderator feedback at part-power conditions.
The [T(Z)]COLR functions quantify these effects for the range of power shapes, control rod insertion, and power levels characteristic of the operating space. Multiplying [T(Z)]COLR by the measured full power, unrodded FXYM(Z) value, and the factor that accounts for manufacturing and measurement uncertainties gives FQW(Z), the maximum total peaking factor postulated for non-equilibrium RAOC operation.
The limit with which FQW(Z) is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-13 Revision 98 BASES SURVEILLANCE SR 3.2.1.2 (continued)
REQUIREMENTS The [T(Z)]COLR functions are specified in the COLR for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations.
FQW(Z) evaluations are not applicable for axial core regions, measured in percent of core height:
- a.
Lower core region, from 0 to 15% inclusive,
- b.
Upper core region, from 85 to 100% inclusive,
- c.
Grid plane regions, + 2% inclusive, and
- d.
Core plane regions, within + 2% of the bank demand position of the control banks.
These regions of the core are excluded from the evaluation because of the low probability that they would be more limiting in the safety analyses and because of the difficulty of making a precise measurement in these regions. The excluded regions at the top and bottom of the core are specified in the COLR and are defined to ensure that the minimum margin location is adequately surveilled. A slightly smaller exclusion zone may be specified, if necessary, to include the limiting margin location in the surveilled region of the core.
SR 3.2.1.2 requires a Surveillance of FQW(Z) during the initial startup following each refueling within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding 75% RTP.
THERMAL POWER levels below 75% are typically non-limiting with respect to the limit for FQW(Z). Furthermore, startup physics testing and flux symmetry measurements, also performed at low power, provide confirmation that the core is operating as expected. This Frequency ensures that verification of FQW(Z) is performed prior to extended operation at power levels where the maximum permitted peak LHR could be challenged and that the first required performance of SR 3.2.1.2 after a refueling is performed at a power level high enough to provide a high level of confidence in the accuracy of the Surveillance result.
Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions required to perform the Surveillance.
If a previous Surveillance of FQW(Z) was performed at part power conditions, SR 3.2.1.2 also requires the FQW(Z) be verified at power levels 10% RTP above the THERMAL POWER of its last verification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions. This ensures that FQW(Z) is within its limit using radial peaking factors measured at the higher power level.
FQ(Z) (RAOC - T(Z) Methodology)
B 3.2.1 Wolf Creek - Unit 1 B 3.2.1-14 Revision 98 BASES SURVEILLANCE SR 3.2.1.2 (continued)
REQUIREMENTS The allowance of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions will provide a more accurate measurement of FQW(Z) by allowing sufficient time to achieve equilibrium conditions and obtain the power distribution measurement.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES
- 1.
10 CFR 50.46, 1974.
- 2.
USAR, Section 15.4.8.
- 3.
10 CFR 50, Appendix A, GDC 26.
- 4.
WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.
- 5.
Performance Improvement Request 2005-3311.
- 6.
WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994 (including Addendum 4, September 2012).
- 7.
WCAP-17661-P-A, Revision 1, Improved RAOC and CAOC FQ Surveillance Technical Specifications, February 2019.
RCS Leakage Detection Instrumentation B 3.4.15 Wolf Creek - Unit 1 B 3.4.15-3 Revision 98 BASES APPLICABLE damage RCS supports, core cooling equipment or core internals. This SAFETY ANALYSES concern was first identified as Multiplant Action (MPA) D-10 and (continued) subsequently as Unresolved Safety Issue (USI) A-2, Asymmetric LOCA Loads. This issue was discussed in Reference 4.
The resolution of USI A-2 for Westinghouse PWRs was the use of fracture mechanics technology for RCS piping > 10 inches diameter (Ref.
5). This technology became known as leak-before-break (LBB). Included within the LBB methodology was the requirement to have leak detection systems capable of detecting a 1.0 gpm leak within four hours. This leakage rate is designed to ensure that adequate margins exist to detect leaks in a timely manner during normal operating conditions.
The need to evaluate the severity of an alarm or an indication is important to the operators, and the ability to compare and verify with indications from other systems is necessary. The system response times and sensitivities are described in the USAR (Ref. 3). Multiple instrument locations are utilized, if needed, to ensure that the transport delay time of the leakage from its source to an instrument location yields an acceptable overall response time.
The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring RCS LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE provides quantitative information to the operators, allowing them to take corrective action should a leak occur detrimental to the safety of the unit and the public.
RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2)(ii).
LCO One method of protecting against large RCS leakage derives from the ability of instruments to rapidly detect extremely small leaks. This LCO requires instruments of diverse monitoring principles to be OPERABLE to provide a high degree of confidence that extremely small leaks are detected in time to allow actions to place the plant in a safe condition, when RCS LEAKAGE indicates possible RCPB degradation.
The LCO is satisfied when monitors of diverse measurement means are available. Thus, the Containment Sump Level and Flow Monitoring System, one containment atmosphere particulate radioactivity monitor and the Containment Cooler Condensate Flow Monitoring System provide an acceptable minimum.
The Containment Sump Level and Flow Monitoring System is considered OPERABLE when it is capable of collecting, measuring and transmitting RCS LEAKAGE to the installed control room indication. The plant
RCS Leakage Detection Instrumentation B 3.4.15 Wolf Creek - Unit 1 B 3.4.15-4 Revision 98 BASES LCO process computer calculates leakage rate from increases in sump level.
(continued)
The leak rate may also be calculated manually once per shift during periods when the plant process computer is out of service.
For containment atmosphere particulate radioactivity monitor instrumentation, OPERABILITY involves more than OPERABILITY of the channel electronics. OPERABILITY also requires correct valve lineups, sample pump operation, sample line insulation and heat tracing, as well as detector OPERABILITY, since these supporting features are necessary for the monitors to rapidly detect RCS LEAKAGE.
The Containment Cooler Condensate Flow Monitoring System is considered OPERABLE when it is capable of measuring liquid flow from the containment coolers. An OPERABLE Containment Cooler Condensate Flow Monitoring System consists of a containment cooler drain collection header, a vertical standpipe, valving, and standpipe level indication for each cooler. Additionally, the plant process computer must be capable of calculating the leakage rate indicated by the condensate being collected. At least two containment coolers must be operating for the Containment Cooler Condensate Flow Monitoring System to be OPERABLE (Ref. 6).
Each containment coolers air inlet temperature, relative humidity and dewpoint determines the amount of condensate the cooler will produce.
Historically, the D containment cooler has produced the least amount of condensate. A containment cooler that is producing a small amount of condensate may result in a stable standpipe level and infrequent actuation of the automatic dump valve. Indications that the automatic dump valve may not be operating properly are no level indication in the standpipe or overfilling of the containment cooling drip pan (Ref. 7).
APPLICABILITY Because of elevated RCS temperature and pressure in MODES 1, 2, 3, and 4, RCS leakage detection instrumentation is required to be OPERABLE.
In MODE 5 or 6, the temperature is required to be 200°F and pressure is maintained low or at atmospheric pressure. Since the temperatures and pressures are far lower than those for MODES 1, 2, 3, and 4, the likelihood of leakage and crack propagation are much smaller. Therefore, the requirements of this LCO are not applicable in MODES 5 and 6.
ACTIONS A.1 and A.2 A primary system leak would result in reactor coolant flowing into the containment normal sumps or into the instrument tunnel sump. Indication of increasing sump level is transmitted to the control room by means of
RCS Leakage Detection Instrumentation B 3.4.15 Wolf Creek - Unit 1 B 3.4.15-5 Revision 98 BASES ACTIONS A.1 and A.2 (continued) individual sump level transmitters. This information is used to provide measurement of low leakage by monitoring level increase versus time.
With the required Containment Sump Level and Flow Monitoring System inoperable, no other form of sampling can provide the equivalent information; however, the containment atmosphere particulate radioactivity monitor will provide indications of changes in leakage.
Together with the atmosphere monitor, the periodic surveillance for RCS water inventory balance, SR 3.4.13.1, must be performed at an increased frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to provide information that is adequate to detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation (stable RCS pressure, temperature, power level, pressurizer and makeup tank level, makeup and letdown, and RCP seal injection and return flows).
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.
Restoration of the required Containment Sump Level and Flow Monitoring System to OPERABLE status within a Completion Time of 30 days is required to regain the function after the systems failure. This time is acceptable, considering the Frequency and adequacy of the RCS water inventory balance required by Required Action A.1. The Completion Time is modified by a Note indicating that the 30 days is extended until startup from a plant shutdown or startup from Refueling Outage 20.
B.1.1, B.1.2, B.2.1 and B.2.2 With the containment atmosphere particulate radioactivity monitoring instrumentation channel inoperable, alternative action is required. Either samples of the containment atmosphere must be taken and analyzed for particulate radioactivity or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information.
Alternatively, continued operation is allowed if the containment air cooler condensate monitoring system is OPERABLE, provided grab samples are taken or water inventory balances are performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With a sample obtained and analyzed or water inventory balance performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor may be operated for up to 30 days to allow restoration of the required containment atmosphere particulate radioactivity monitor.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval provides periodic information that is adequate to detect leakage. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation (stable RCS pressure, temperature, power level, pressurizer and makeup tank level, makeup and letdown, and RCP seal injection and return flows).
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established. The 30 day
RCS Leakage Detection Instrumentation B 3.4.15 Wolf Creek - Unit 1 B 3.4.15-6 Revision 98 BASES ACTIONS B.1.1, B.1.2, B.2.1 and B.2.2 (continued)
Completion Time recognizes at least one other form of leakage detection is available.
C.1 and C.2 With the required containment cooler condensate monitoring system inoperable, alternative action is again required. Either SR 3.4.15.1 must be performed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information.
Provided a CHANNEL CHECK is performed every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or a water inventory balance is performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, reactor operation may continue while awaiting restoration of the containment cooler condensate monitoring system to OPERABLE status.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval provides periodic information that is adequate to detect RCS LEAKAGE. A Note is added allowing that SR 3.4.13.1 is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation (stable RCS pressure, temperature, power level, pressurizer and makeup tank level, makeup and letdown, and RCP seal injection and return flows.)
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.
D.1 and D.2 With the required containment atmosphere particulate radioactivity monitor and the required Containment Cooler Condensate Monitoring System inoperable, the means of detecting leakage is the Containment Sump Level and Flow Monitoring System. This Condition does not provide all the required diverse means of leakage detection. The Required Action is to restore either of the inoperable required monitoring methods to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a reduced configuration for a lengthy time period.
Refer to LCO 3.3.6, Containment Purge Isolation Instrumentation, upon a loss of the required containment atmosphere radioactivity monitor to ensure LCO requirements are met.
RCS Leakage Detection Instrumentation B 3.4.15 Wolf Creek - Unit 1 B 3.4.15-7 Revision 98 BASES ACTIONS E.1 and E.2 (continued)
If a Required Action of Condition A, B, C or D cannot be met, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
F.1 With all required monitoring methods inoperable, no automatic means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.
SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere particulate radioactivity monitor. The check gives reasonable confidence that the channel is operating properly.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere particulate radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner.
The test verifies the alarm setpoint and relative accuracy of the instrument string. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.15.3, SR 3.4.15.4, and SR 3.4.15.5 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
RCS Leakage Detection Instrumentation B 3.4.15 Wolf Creek - Unit 1 B 3.4.15-8 Revision 98 BASES REFERENCES
- 1.
10 CFR 50, Appendix A,Section IV, GDC 30.
- 2.
- 3.
USAR, Section 5.2.5.
- 4.
NUREG-609, Asymmetric Blowdown Loads on PWR Primary Systems, 1981.
- 5.
Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops.
- 6.
USAR, Section 6.2.2.2.2.
- 7.
Performance Improvement Request 2005-2823.
- 8.
NRC letter, Wolf Creek Generating Station - License Amendment Request to Change the Reactor Coolant System Leakage Detection Instrumentation Methodology (TAC No. MC8214), May 16, 2006.
AC Sources - Operating B 3.8.1 Wolf Creek - Unit 1 B 3.8.1-1 Revision 98 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources - Operating BASES BACKGROUND The unit Class 1E AC Electrical Power Distribution System AC sources consist of the offsite power sources (preferred power sources, normal and alternate from the redundant Engineered Safety Feature (ESF) transformers), and the onsite standby power sources (Train A and Train B diesel generators (DGs)). As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the ESF systems.
The onsite Class 1E AC Distribution System is divided into redundant load groups (trains) so that the loss of any one group does not prevent the minimum safety functions from being performed. Each train has connections to its preferred offsite power source and a single DG.
Offsite power is supplied to the unit switchyard from the transmission network by four transmission lines. From the switchyard, two electrically and physically separated circuits provide AC power, through the ESF transformers, to the 4.16 kV ESF buses. A detailed description of the offsite power network and the circuits to the Class 1E ESF buses is found in the USAR, Chapter 8 (Ref. 2).
An offsite circuit consists of all breakers, transformers, voltage regulating tap changers (when installed), switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class 1E ESF bus(es).
Certain required unit loads are returned to service in a predetermined sequence in order to prevent overloading the transformer supplying offsite power to the onsite Class 1E Distribution System. Within 1 minute after the initiating signal is received, all automatic and permanently connected loads needed to recover the unit or maintain it in a safe condition are returned to service via the load sequencer.
The onsite standby power source for each 4.16 kV ESF bus is a dedicated DG. DGs A and B are dedicated to ESF buses NB01 and NB02, respectively. The DG starts automatically on a safety injection (SI) signal or on an ESF bus undervoltage signal. A degraded voltage signal produces an undervoltage condition by opening the normal and alternate feeder breakers to the bus(es) experiencing degraded voltage. Both signals are initiated from the load shedder and emergency load sequencer (LSELS). OPERABILITY of the undervoltage and degraded voltage instrumentation functions are addressed in LCO 3.3.5, Loss of