NLS2011093, Submittal of Technical Specification Bases Changes

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Submittal of Technical Specification Bases Changes
ML11291A119
Person / Time
Site: Cooper Entergy icon.png
Issue date: 10/14/2011
From: Vanderkamp D
Nebraska Public Power District (NPPD)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2011093
Download: ML11291A119 (40)
v • d • e


Text

H Nebraska Public Power District Always there when you need us NLS2011093 October 14, 2011 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Technical Specification Bases Changes Cooper Nuclear Station, Docket No. 50-298, DPR-46

Dear Sir or Madam:

The purpose of this letter is to provide changes to the Cooper Nuclear Station (CNS) Technical Specification Bases implemented without prior Nuclear Regulatory Commission approval. In accordance with the requirements of CNS Technical Specification 5.5.1 0.d, these changes are provided on a frequency consistent with 10 CFR 50.71(e). The enclosed Bases changes are for the time period from March 5, 2010, through September 16, 2011. Also enclosed are filing instructions and an updated List of Effective Pages for the CNS Technical Specification Bases.

If you have any questions regarding this submittal, please contact me at (402) 825-2904.

Sincerely, 0414dJ vall 4 David W. Van Der Kamp Licensing Manager

/lb Enclosure cc: Regional Administrator, w/enclosure USNRC - Region IV Cooper Project Manager, w/enclosure USNRC - NRR Project Directorate IV-1 Senior Resident Inspector, w/enclosure (per controlled document distribution)

USNRC - CNS NPG Distribution, w/o enclosure CNS Records, w/enclosure /k*-*

COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com

4 ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS@4 Correspondence Number: NLS2011093 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE None 4 4 4 4 4 4 4 4 4 4 4 4

[ PROCEDURE 0.42 REVISION 27 PAGE 18 OF 25

NLS2011093 ENCLOSURE TECHNICAL SPECIFICATION BASES CHANGES

FILING INSTRUCTIONS TECHNICAL SPECIFICATION BASES REMOVE INSERT List of Effective Pages - Bases 1 through 8 1 through 8 Bases Pages B 3.3-45 B 3.3-45 B 3.3-82 B 3.3-82 B 3.3-115 B 3.3-115 B 3.3-148 B 3.3-148 B 3.3-170 B 3.3-170 B 3.4-17 B 3.4-17 B 3.4-18 B 3.4-18 B 3.5-12 B 3.5-12 B 3.6-37 B 3.6-37 B 3.6-38 B 3.6-38 B 3.6-63 B 3.6-63 B 3.7-10 B 3.7-10 B 3.8-1 B 3.8-1 B 3.8-20 B 3.8-20 B 3.8-21 B 3.8-21 B 3.8-22 B 3.8-22 B 3.8-23 B 3.8-23 B 3.8-24 B 3.8-24 B 3.8-25 B 3.8-25 B 3.8-36 B 3.8-36 B 3.8-41 B 3.8-41 B 3.8-43 B 3.8-43 B 3.8-46 B 3.8-46 B 3.8-47 B 3.8-47 B 3.8-48 B 3.8-48 B 3.8-49 B 3.8-49 B 3.8-50 B 3.8-50 B 3.8-51 B 3.8-51 LIST OF EFFECTIVE PAGES - BASES Page No. Revision No./Date Page No. Revision No./Date ii 0 B 3.1-13 12/18/03 iii 0 B 3.1-14 6/10/99 0 B 3.1-15 6/10/99 B 3.1-16 12/03/09 B 2.0-1 12/18/03 B 3.1-17 6/10/99 B 2.0-2 0 B 3.1-18 07/16/08 B 2.0-3 0 B 3.1-19 12/03/09 B 2.0-4 6/10/99 B 3.1-20 12/03/09 B 2.0-5 09/25/09 B 3.1-21 12/18/03 B 2.0-6 09/25/09 B 3.1-22 0 B 2.0-7 0 B 3.1-23 0 B 2.0-8 09/25/09 B 3.1-24 0 B 3.1-25 05/09/06 B 3.0-1 06/30/06 B 3.1-26 02/02/06 B 3.0-2 0 B 3.1-27 05/09/06 B 3.0-3 0 B 3.1-28 12/18/03 B 3.0-4 0 B 3.1-29 0 B 3.0-5 09/18/09 B 3.1-30 0 B 3.0-6 09/18/09 B 3.1-31 0 B 3.0-7 09/18/09 B 3.1-32 0 B 3.0-8 09/18/09 B 3.1-33 01/30/03 B 3.0-9 09/18/09 B 3.1-34 07/16/08

( B 3.0-10 09/18/09 B 3.1-35 07/16/08 B 3.0-11 09/18/09 B 3.1-36 07/16/08 B 3.0-12 09/18/09 B 3.1-37 07/16/08 B 3.0-13 09/18/09 B 3.1-38 07/16/08 B 3.0-14 09/18/09 B 3.1-39 09/25/09 B 3.0-15 09/18/09 B 3.1-40 09/25/09 B 3.0-16 09/18/09 B 3.1-41 09125109 B 3.0-17 09/18/09 B 3.1-42 09/25/09 B 3.0-18 09/18/09 B 3.1-43 09/25/09 B 3.1-44 09/25/09 B 3.1-1 6/10/99 B 3.1-45 09/25/09 B 3.1-2 6/10/99 B 3.1-46 09/25/09 B 3.1-3 6/10/99 B 3.1-47 09/25/09 B 3.1-4 6/10/99 B 3.1-48 0 B 3.1-5 6/10/99 B 3.1-49 0 B 3.1-6 6/10/99 B 3.1-50 6/10/99 B 3.1-7 12/18/03 B 3.1-51 09/25/09 B 3.1-8 12/18/03 B 3.1-9 6/10/99 B 3.2-1 01/27/06 B 3.1-10 6/10/99 B 3.2-2 6/10/99 B 3.1-11 6/10/99 B 3.2-3 6/10/99 B 3.1-12 12/18/03 B 3.2-4 4/12/00 Cooper 1 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.2-5 6/10/99 B 3.3-38 6/28101 B 3.2-6 0 B 3.3-39 6/28/01 B 3.2-7 0 B 3.3-40 11/04/06 B 3.2-8 0 B 3.3-41 6/28/01 B 3.2-9 0 B 3.3-42 0 B 3.2-10 4/12/00 B 3.3-43 07/16/08 B 3.3-44 0 B 3.3-1 0 B 3.3-45 08/17/11 B 3.3-2 0 B 3.3-46 07/16/08 B 3.3-3 0 B 3.3-47 1/14/05 B 3.3-4 1 B 3.3-48 1/14/05 B 3.3-5 1 B 3.3-49 1/14/05 B 3.3-6 1 B 3.3-50 07/16/08 B 3.3-7 8/29/02 B 3.3-51 07/16/08 B 3.3-8 6/7/02 B 3.3-52 07/16/08 B 3.3-9 6/7/02 B 3.3-53 02/20/07 B 3.3-10 6/7/02 B 3.3-54 1/14/05 B 3.3-11 1 B 3.3-55 0 B 3.3-12 1 B 3.3-56 1 B 3.3-13 1 B 3.3-57 1 B 3.3-14 1 B 3.3-58 1 B 3.3-15 1 B 3.3-59 0 B 3.3-16 6/28/01 B 3.3-60 0 B 3.3-17 08/28/08 B 3.3-61 02/20/07 B 3.3-18 07/16/08 B 3.3-62 0 B 3.3-19 07/16/08 B 3.3-63 1 B 3.3-20 6/28/01 B 3.3-64 08/08/08 B 3.3-21 6/28/01 B 3.3-65 08/08/08 B 3.3-22 6/28/01 B 3.3-66 1 B 3.3-23 6/28/01 B 3.3-67 0 B 3.3-24 6/28/01 B 3.3-68 08/01/07 B 3.3-25 6/28/01 B 3.3-69 09/18/09 B 3.3-26 6/28/01 B 3.3-70 0 B 3.3-27 6/28/01 B 3.3-71 0 B 3.3-28 05/05/06 B 3.3-72 1 B 3.3-29 02/20/07 B 3.3-73 1 B 3.3-30 02/20/07 B 3.3-74 0 B 3.3-31 07/16/08 B 3.3-75 1 B 3.3-32 6/28/01 B 3.3-76 09/18/09 B 3.3-33 0 B 3.3-77 0 B 3.3-34 0 B 3.3-78 06/07/06 B 3.3-35 0 B 3.3-79 1 B 3.3-36 0 B 3.3-80 0 B 3.3-37 0 B 3.3-81 0 Cooper 2 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.3-82 03/25/11 B 3.3-126 0 B 3.3-83 0 B 3.3-127 0 B 3.3-84 0 B 3.3-128 6/10/99 B 3.3-85 0 B 3.3-129 1 B 3.3-86 0 B 3.3-130 0 B 3.3-87 0 B 3.3-131 0 B 3.3-88 6/28/01 B 3.3-132 0 B 3.3-89 02/20/07 B 3.3-133 0 B 3.3-90 0 B 3.3-134 0 B 3.3-91 0 B 3.3-135 6/28/01 B 3.3-92 0 B 3.3-136 02/20/07 B 3.3-93 1 B 3.3-137 0 B 3.3-94 1 B 3.3-138 01/24/06 B 3.3-95 0 B 3.3-139 01/24/06 B 3.3-96 05/09/06 B 3.3-140 01/24/06 B 3.3-97 0 B 3.3-141 01/24/06 B 3.3-98 05/09/06 B 3.3-142 01/24/06 B 3.3-99 0 B 3.3-143 0 B 3.3-100 05/09/06 B 3.3-144 09/25/09 B 3.3-101 05/09/06 B 3.3-145 0 B 3.3-102 1 B 3.3-146 0 B 3.3-103 6/10/99 B 3.3-147 0 B 3.3-104 05/09/06 B 3.3-148 02/22/11 B 3.3-105 1 B 3.3-149 12/22/05 B 3.3-106 0 B 3.3-150 4/12/00 B 3.3-107 0 B 3.3-151 09/25/09 B 3.3-108 8/29/02 B 3.3-152 4/12/00 B 3.3-109 0 B 3.3-153 0 B 3.3-110 0 B 3.3-154 4/19/05 B 3.3-111 0 B 3.3-155 02/16/10 B 3.3-112 0 B 3.3-156 12/22/05 B 3.3-113 1 B 3.3-157 12/22/05 B 3.3-114 0 B 3.3-158 0 B 3.3-115 02/22/11 B 3.3-159 0 B 3.3-116 0 B 3.3-160 0 B 3.3-117 0 B 3.3-161 0 B 3.3-118 0 B 3.3-162 0 B 3.3-119 0 B 3.3-163 6/28/01 B 3.3-120 0 B 3.3-164 6/28/01 B 3.3-121 0 B 3.3-165 02/20/07 B 3.3-122 0 B 3.3-166 6/28/01 B 3.3-123 6/28/01 B 3.3-167 01/24/06 B 3.3-124 02/20/07 B 3.3-168 01/24/06 B 3.3-125 6/28/01 B 3.3-169 2/10/05 Cooper 3 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.3-170 02/22/11 B 3.4-3 0 B 3.3-171 12/22/05 B 3.4-4 0 B 3.3-172 10/05/06 B 3.4-5 0 B 3.3-173 0 B 3.4-6 0 B 3.3-174 0 B 3.4-7 4/12/00 B 3.3-175 6/28/01 B 3.4-8 0 B 3.3-176 02/20/07 B 3.4-9 0 B 3.3-177 6/28/01 B 3.4-10 0 B 3.3-178 0 B 3.4-11 1 B 3.3-179 0 B 3.4-12 1 B 3.3-180 0 B 3.4-13 4/12/00 B 3.3-181 0 B 3.4-14 0 B 3.3-182 6/28/01 B 3.4-15 0 B 3.3-183 02/20/07 B 3.4-16 0 B 3.3-184 6/28/01 B 3.4-17 04/28/10 B 3.3-185 09/25/09 B 3.4-18 04/28/10 B 3.3-186 11/04/01 B 3.4-19 0 B 3.3-187 12/22/05 B 3.4-20 0 B 3.3-188 12/22/05 B 3.4-21 0 B 3.3-189 10/05/06 B 3.4-22 0 B 3.3-190 10/05/06 B 3.4-23 0 B 3.3-191 10/05/06 B 3.4-24 0 B 3.3-192 10/05/06 B 3.4-25 0 B 3.3-193 02/20/07 B 3.4-26 09/18/09 B 3.3-194 11/04/01 B 3.4-27 09/18/09 B 3.3-195 11/04/01 B 3.4-28 6/28/01 B 3.3-196 11/04/01 B 3.4-29 09/25/09 B 3.3-197 11/04/01 B 3.4-30 0 B 3.3-198 11/04/01 B 3.4-31 09/18/09 B 3.3-199 11/04/01 B 3.4-32 09/25/09 B 3.3-200 11/04/01 B 3.4-33 1 B 3.3-201 11/04/01 B 3.4-34 0 B 3.3-202 11/04/01 B 3.4-35 09/18/09 B 3.3-203 11/04/01 B 3.4-36 0 B 3.3-204 02/20/07 B 3.4-37 0 B 3.3-205 11/04/01 B 3.4-38 0 B 3.3-206 11/04/01 B 3.4-39 1 B 3.3-207 11/04/01 B 3.4-40 0 B 3.3-208 11/04/01 B 3.4-41 0 B 3.3-209 11/04/01 B 3.4-42 0 B 3.3-210 02/20/07 B 3.4-43 0 B 3.4-44 08/11/04 B 3.4-1 0 B 3.4-45 08/11/04 B 3.4-2 0 B 3.4-46 04/11/06 Cooper 4 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.4-47 0 B 3.6-4 11/06/06 B 3.4-48 0 B 3.6-5 09/30/08 B 3.4-49 08/11/04 B 3.6-6 0 B 3.4-50 04/11/06 B 3.6-7 09/30/08 B 3.4-51 0 B 3.6-8 0 B 3.4-52 08/11/04 B 3.6-9 0 B 3.4-53 0 B 3.6-10 0 B 3.4-54 0 B 3.6-11 0 B 3.4-55 0 B 3.6-12 3/8/00 B 3.6-13 09/30/08 B 3.5-1 1 B 3.6-14 3/8/00 B 3.5-2 11/24/03 B 3.6-15 0 B 3.5-3 0 B 3.6-16 1 B 3.5-4 0 B 3.6-17 0 B 3.5-5 04/26/04 B 3.6-18 0 B 3.5-6 09/18/09 B 3.6-19 11/28/01 B 3.5-7 04/26/04 B 3.6-20 11/28/01 B 3.5-8 04/26/04 B 3.6-21 11/28/01 B 3.5-9 1 B 3.6-22 11/28/01 B 3.5-10 0 B 3.6-23 11/28/01 B 3.5-11 0 B 3.6-24 1 B 3.5-12 04/28/10 B 3.6-25 1 B 3.5-13 4/19/00 B 3.6-26 3/8/00 B 3.5-14 02/20/07 B 3.6-27 11/04/01 B 3.5-15 02/20/07 B 3.6-28 09/25/09 B 3.5-16 0 B 3.6-29 09/25/09 B 3.5-17 11/23/99 B 3.6-30 09/30/08 B 3.5-18 12/18/03 B 3.6-31 09/30/08 B 3.5-19 0 B 3.6-32 12/27/02 B 3.5-20 0 B 3.6-33 12/27/02 B 3.5-21 0 B 3.6-34 12/14/01 B 3.5-22 0 B 3.6-35 0 B 3.5-23 12/18/03 B 3.6-36 0 B 3.5-24 0 B 3.6-37 04/28/10 B 3.5-25 1 B 3.6-38 04/28/10 B 3.5-26 09/18/09 B 3.6-39 0 B 3.5-27 09/18/09 B 3.6-40 0 B 3.5-28 09/18/09 B 3.6-41 0 B 3.5-29 02/20/07 B 3.6-42 0 B 3.5-30 12/18/03 B 3.6-43 0 B 3.6-44 02/20/07 B 3.6-1 3/8/00 B 3.6-45 0 B 3.6-2 09/30/08 B 3.6-46 6/10/99 B 3.6-3 3/8/00 B 3.6-47 0

( Cooper 5 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.6-48 0 B 3.7-7 05/06/09 B 3.6-49 0 B 3.7-8 05/06/09 B 3.6-50 6/10/99 B 3.7-9 05/06/09 B 3.6-51 0 B 3.7-10 02/22/11 B 3.6-52 8/13/02 B 3.7-11 05/06/09 B 3.6-53 0 B 3.7-12 05/06/09 B 3.6-54 0 B 3.7-13 05/06/09 B 3.6-55 8/13/02 B 3.7-14 05/06/09 B 3.6-56 8/13/02 B 3.7-15 05/06/09 B 3.6-57 0 B 3.7-16 05/06/09 B 3.6-58 0 B 3.7-17 10/29/08 B 3.6-59 0 B 3.7-18 10/29/08 B 3.6-60 0 B 3.7-19 10/29/08 B 3.6-61 0 B 3.7-20 10/29/08 B 3.6-62 0 B 3.7-21 10/29/08 B 3.6-63 04/28/10 B 3.7-22 10/29/08 B 3.6-64 0 B 3.7-23 10/29/08 B 3.6-65 0 B 3.7-24 10/29/08 B 3.6-66 0 B 3.7-25 04/30/09 B 3.6-67 10/05/06 B 3.7-26 10/29/08 B 3.6-68 10/05/06 B 3.7-27 10/29/08 B 3.6-69 10/05/06 B 3.7-28 10/29/08 B 3.6-70 10/05/06 B 3.7-29 10/29/08 B 3.6-71 0 B 3.7-30 10/29/08 B 3.6-72 10/05/06 B 3.7-31 10/29/08 B 3.6-73 10/05/06 B 3.7-32 10/29/08 B 3.6-74 10/05/06 B 3.7-33 10/29/08 B 3.6-75 3/8/00 B 3.7-34 10/29/08 B 3.6-76 10/05/06 B 3.6-77 12/18/03 B 3.8-1 02/22/11 B 3.6-78 11/04/01 B 3.8-2 4/16/02 B 3.6-79 12/18/03 B 3.8-3 3/15/01 B 3.6-80 10/05/06 B 3.8-4 3/15/01 B 3.6-81 10/05/06 B 3.8-5 4/16/02 B 3.6-82 10/05/06 B 3.8-6 09/18/09 B 3.6-83 10/05/06 B 3.8-7 09/18/09 B 3.6-84 12/18/03 B 3.8-8 09/18/09 B 3.8-9 1/17/05 B 3.7-1 1 B 3.8-10 07/01/04 B 3.7-2 0 B 3.8-11 12/22/05 B 3.7-3 03/24/04 B 3.8-12 12/22/05 B 3.7-4 0 B 3.8-13 1 B 3.7-5 0 B 3.8-14 1 B 3.7-6 05/06/09 B 3.8-15 12/18/03 Cooper 6 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.8-16 1 B 3.8-60 0 B 3.8-17 0 B 3.8-61 0 B 3.8-18 0 B 3.8-62 0 B 3.8-19 04/30/09 B 3.8-63 0 B 3.8-20 04/28/10 B 3.8-64 0 B 3.8-21 03/25/11 B 3.8-65 0 B 3.8-22 03/25/11 B 3.8-66 0 B 3.8-23 03/25/11 B 3.8-67 10/14/04 B 3.8-24 03/25/11 B 3.8-68 0 B 3.8-25 04/28/10 B 3.8-69 10/14/04 B 3.8-26 0 B 3.8-70 04/11/06 B 3.8-27 0 B 3.8-71 0 B 3.8-28 05/09/06 B 3.8-72 0 B 3.8-29 0 B 3.8-73 0 B 3.8-30 0 B 3.8-74 0 B 3.8-31 05/09/06 B 3.8-75 0 B 3.8-32 0 B 3.8-33 12/22/05 B 3.9-1 12/18/03 B 3.8-34 12/22/05 B 3.9-2 0 B 3.8-35 10/21/04 B 3.9-3 05/09/06 B 3.8-36 08/23/11 B 3.9-4 05/09/06 B 3.8-37 1 B 3.9-5 05/09/06 B 3.8-38 0 B 3.9-6 05/09/06 B 3.8-39 0 B 3.9-7 05/09/06 B 3.8-40 1 B 3.9-8 05/09/06 B 3.8-41 08/23/11 B 3.9-9 12/18/03 B 3.8-42 0 B 3.9-10 0 B 3.8-43 04/22/10 B 3.9-11 12/18/03 B 3.8-44 0 B 3.9-12 12/18/03 B 3.8-45 04/11/06 B 3.9-13 0 B 3.8-46 04/22/10 B 3.9-14 0 B 3.8-47 04/22/10 B 3.9-15 12/18/03 B 3.8-48 04/22/10 B 3.9-16 12/18/03 B 3.8-49 04/22/10 B 3.9-17 0 B 3.8-50 04/22/10 B 3.9-18 12/18/03 B 3.8-51 04/22/10 B 3.9-19 10/05/06 B 3.8-52 0 B 3.9-20 0 B 3.8-53 0 B 3.9-21 10/05/06 B 3.8-54 0 B 3.9-22 0 B 3.8-55 0 B 3.9-23 0 B 3.8-56 0 B 3.9-24 0 B 3.8-57 0 B 3.9-25 0 B 3.8-58 0 B 3.9-26 0 B 3.8-59 0 B 3.9-27 0 Cooper 7 08/23/11

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No. Revision No./Date Page No. Revision No./Date B 3.9-28 0 B 3.9-29 0 B 3.9-30 0 B 3.10-1 11/06/06 B 3.10-2 11/06/06 B 3.10-3 11/06/06 B 3.10-4 11/06/06 B 3.10-5 11/06/06 B 3.10-6 0 B 3.10-7 0 B 3.10-8 0 B 3.10-9 0 B 3.10-10 0 B 3.10-11 0 B 3.10-12 0 B 3.10-13 0 B 3.10-14 0 B 3.10-15 0 B 3.10-16 0 B 3.10-17 0 B 3.10-18 0 B 3.10-19 0 B 3.10-20 0 B 3.10-21 0 B 3.10-22 0 B 3.10-23 0 B 3.10-24 0 B 3.10-25 0 B 3.10-26 6/10/99 B 3.10-27 6/10/99 B 3.10-28 0 B 3.10-29 6/10/99 B 3.10-30 0 B 3.10-31 07/16/08 B 3.10-32 0 B 3.10-33 0 B 3.10-34 0 B 3.10-35 0 B 3.10-36 0 B 3.10-37 0 B 3.10-38 0 B 3.10-39 0 08/23/11 8

Cooper 8 08/23/11

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Nominal trip setpoints are specified in the setpoint calculations. The setpoint calculations are performed using methodology described in NEDC-31336P-A, "General Electric Instrument Setpoint Methodology,"

dated September 1996. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors.

The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The RBM is assumed to mitigate the consequences of an RWE event when operating > 30% RTP (analytical limit) and a peripheral control rod is not selected. Below this power level or if a peripheral control rod is selected, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE (Ref. 3).

When operating < 90% RTP, analyses (Ref. 3) have shown that with an initial MCPR > 1.70, no RWE event will result in exceeding the MCPR SL.

Also, the analyses demonstrate that when operating at > 90% RTP with MCPR > 1.40, no RWE event will result in exceeding the MCPR SL (Ref. 3). Therefore, under these conditions, the RBM is also not required to be OPERABLE.

2. Rod Worth Minimizer The RWM is a backup to operator control of the rod sequences. The RWM enforces the banked position withdrawal sequence (BPWS) by alerting the operator when the rod pattern is not in accordance with BPWS. Compliance with BPWS ensures that the initial conditions of the CRDA analysis are not violated.

Cooper B 3.3-45 08/17/11

ATWS-RPT Instrumentation B 3.3.4.1 BASES BACKGROUND (continued)

Vessel Water Level-Low Low (Level 2) channels or two Reactor Pressure-High channels) will actuate one of the trip coils in each RRMG field breaker, thus tripping both recirculation pumps.

APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY The ATWS-RPT initiates an RPT to aid in preserving the integrity of the fuel cladding following events in which a scram does not, but should, occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of 10 CFR 50.36(c)(2)(ii)

(Ref. 2).

The OPERABILITY of the ATWS-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1.2. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated RRMG field breakers.

Allowable Values are specified for each ATWS-RPT Function specified in the LCO. Nominal trip setpoints are specified in the setpoint calculations.

The setpoint calculations are performed using methodology described in NEDC-31336P-A, "General Electric Instrument Setpoint Methodology,"

dated September 1996. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., switch) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and (continued)

Cooper B 3.3-82 03/25/11

ECCS Instrumentation B 3.3.5.1 BASES ACTIONS (continued)

C.1 and C.2 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function result in redundant automatic initiation capability being lost for the feature(s).

Required Action C.1 features would be those that are initiated by Functions 1.c, 1.e, 2.c, 2.d, and 2.f (i.e., low pressure ECCS).

Redundant automatic initiation capability is lost if either (a) two Function 1.c channels are inoperable such that both trip systems lose initiation capability, (b) two Function i.e channels are inoperable, (c) two Function 2.c channels are inoperable such that both trip systems lose initiation capability, (d) two Function 2.d channels are inoperable such that both trip systems lose initiation capability, or (e) two or more Function 2.f channels are inoperable. In this situation (loss of redundant automatic initiation capability), the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system to be declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g., both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being concurrently declared inoperable. For Functions 1.c, 1.e, 2.d, and 2.f, the affected portions are the associated low pressure ECCS pumps. As noted (Note 1), Required Action C.1 is only applicable in MODES 1, 2, and 3. In MODES 4 and 5, the specific initiation time of the ECCS is not assumed and the probability of a LOCA is lower. Thus, a total loss of automatic initiation capability for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (as allowed by Required Action C.2) is allowed during MODES 4 and 5.

Note 2 states that Required Action C.1 is only applicable for Functions 1.c, 1.e, 2.c, 2.d, and 2.f. Required Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 8 and considered acceptable for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed by Required Action C.2.

Cooper B 3.3-115 02/22/11

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

Primary Containment Isolation 2.a. Reactor Vessel Water Level - Low (Level 3)

Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products.

The isolation of the primary containment on Level 3 supports actions to ensure that offsite dose limits of 10 CFR 50.67 are not exceeded. The Reactor Vessel Water Level - Low (Level 3) Function associated with isolation is implicitly assumed in the USAR analysis as these leakage paths are assumed to be isolated post LOCA.

Reactor Vessel Water Level - Low (Level 3) signals are initiated from four vessel level instrument switches that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level - Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low (Level 3) Allowable Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown.

This Function isolates the Group 2 valves listed in Reference 1.

2.b. Drywell Pressure - High High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 50.67 are not exceeded. The Drywell Pressure- High Function, associated with isolation of the primary containment, is implicitly assumed in the USAR accident analysis as these leakage paths are assumed to be isolated post LOCA.

Cooper B 3.3-148 02/22/11

Secondary Containment Isolation Instrumentation B 3.3.6.2 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level - Low Low (Level 2) Function are available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude the isolation function.

The Reactor Vessel Water Level - Low Low (Level 2) Allowable Value was chosen to be the same as the High Pressure Coolant Injection/

Reactor Core Isolation Cooling (HPCI/RCIC) Reactor Vessel Water Level Low Low (Level 2) Allowable Value (LCO 3.3.5.1 and LCO 3.3.5.2) since this could indicate that the capability to cool the fuel is being threatened).

The Reactor Vessel Water Level - Low Low (Level 2) Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required. In addition, the Function is also required to be OPERABLE during operations with a potential for draining the reactor vessel (OPDRVs) because the capability of isolating potential sources of leakage must be provided to ensure that offsite dose limits are not exceeded if core damage occurs.

This function isolates the Group 6 valves listed in Reference 1.

2. Drywell Pressure-High High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. The Drywell Pressure - High Function associated with isolation is not assumed in any USAR accident or transient analyses, but will provide an isolation and initiation signal. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.

Cooper B 3.3-170 02/22/11

SRVs and SVs B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.3.2 A manual actuation of each SRV (until the main turbine bypass valves have closed to compensate for SRV opening) is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can also be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure and steam flow when the SRVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test.

Adequate pressure at which this test is to be performed is >_500 psig, consistent with the recommendations of the vendor. Adequate steam flow is represented by turbine bypass valves at least 30% open, or total steam flow > 106 lb/hr. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure and steam flow are reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is not considered inoperable.

The 18 month Frequency was developed based on the SRV tests required by the ASME Code for Operation and Maintenance of Nuclear Power Plants (Ref. 6). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Cooper B 3.4-17 04/28/10

SRVs and SVs B 3.4.3 BASES REFERENCES 1. ASME Boiler and Pressure Vessel Code,Section III.

2. USAR, Section IV-4.9.
3. NEDC-31628P, SRV Setpoint Tolerance Analysis for Cooper Nuclear Station, October 1988.
4. USAR,Section XIV.
5. 10 CFR 50.36(c)(2)(ii).
6. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Cooper B 3.4-18 04/28/10

ECCS - Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued) the LPCI subsystem. Acceptable methods of de-energizing the valve include de-energizing breaker control power, racking out the breaker or removing the breaker.

The specified Frequency is once during reactor startup before THERMAL POWER is > 25% RTP. However, this SR is modified by a Note that states the Surveillance is only required to be performed if the last performance was more than 31 days ago. Therefore, implementation of this Note requires this test to be performed during reactor startup before exceeding 25% RTP. Verification during reactor startup prior to reaching

> 25% RTP is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the demonstrated reliability of these valves. If the valve is inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.

SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 7). This periodic Surveillance is performed (in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 8. The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA.

The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow against a system head corresponding to reactor pressure is tested at both the higher and lower operating ranges of the system. The required system head Cooper B 3.5-12 04/28/10

LLS Valves B 3.6.1.6 BASES ACTIONS (continued)

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

SURVEILLANCE REQUIREMENTS SR 3.6.1.6.1 A manual actuation of each LLS valve is performed to verify that the valve and solenoids are functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control or bypass valve, by a change in the measured steam flow, or by any other method that is suitable to verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Adequate pressure at which this test is to be performed is > 500 psig (consistent with the recommendations of the vendor). Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the LLS valves divert steam flow upon opening. Adequate steam flow is represented by turbine bypass valves at least 30% open, or total steam flow > 106 lb/hr. The 18 month Frequency was based on the SRV tests required by the ASME Code for Operation and Maintenance of Nuclear Power Plants (Ref. 3). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Since steam pressure is required to perform the Surveillance, however, and steam may not be available during a unit outage, the Surveillance may be performed during the startup following a unit outage. Unit startup is allowed prior to performing the test because valve OPERABILITY and the setpoints for overpressure protection are verified by Reference 3 prior to valve installation. After adequate reactor steam dome pressure and flow are reached, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed to prepare for and perform the test.

Cooper B 3.6-37 04/28/10

LLS Valves B 3.6.1.6 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.6.2 The LLS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.6.3, "Low-Low Set (LLS) Instrumentation," overlaps this SR to provide complete testing of the safety function.

The 18 month Frequency is based on the need to perform some of the surveillance procedures which satisfy this SR under the conditions that apply during a plant outage and the potential for an unplanned transient if those particular procedures were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation. This prevents a reactor pressure vessel pressure blowdown.

REFERENCES 1. 10 CFR 50.36(c)(2)(ii).

2. NEDE-22197, Safety Relief Valve Low Low Set System and Lower MSIV Water Level Trip for Cooper Nuclear Station, Unit 1, December 1982.
3. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Cooper B 3.6-38 04/28/10

RHR Suppression Pool Cooling B 3.6.2.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.2.3.1 (continued) position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the system is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience.

SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate > 7700 gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASME Code (Ref. 4). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.

REFERENCES 1. USAR, Section XIV-6.

2. 10 CFR 36(c)(2)(ii).
3. NEDC 94-034B, C & D
4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

Cooper B 3.6-63 04/28/10

SW System and UHS B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.7.2.4 This SR verifies that the automatic isolation valves of the SW System will automatically switch to the safety or emergency position to provide cooling water exclusively to the safety related equipment during an accident event. This is demonstrated by the use of an actual or simulated initiation signal. The initiation signal is caused by low SW header pressure (approximately 20 psig). This SR also verifies the automatic start capability of one of the two SW pumps in each subsystem.

Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency. Therefore, this Frequency is concluded to be acceptable from a reliability standpoint.

REFERENCES 1. NEDC 94-255, "Hydraulic Evaluation of Opening in Intake Structure Guide Wall," June 14,1995.

2. USAR, Chapter V.
3. USAR, Chapter XIV.
4. 10 CFR 50.36(c)(2)(ii).
5. NEDC 00-095E, "CNS Reactor Building Post-LOCA Heating Analysis," May 28, 2010.

Cooper B 3.7-10 02/22/11

AC Sources - Operating B 3.8.1 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources - Operating BASES BACKGROUND The unit AC Sources for the Class 1 E AC Electrical Power Distribution System consist of the offsite power sources (preferred power sources, normal and alternates), and the onsite standby power sources (diesel generators (DGs)). As summarized in the USAR, (Ref. 1), the design of the AC electrical power system provides independence and redundancy to ensure an available source of power to the Engineered Safety Feature (ESF) systems.

The Class 1 E AC distribution system is divided into redundant load groups, so loss of any one group does not prevent the minimum safety functions from being performed. Each load group has connections to two qualified offsite power supplies and a single DG.

The offsite power sources are a startup station service transformer (SSST) which connects to the 161 kV switchyard and a separate emergency station service transformer (ESST) energized by a 69 kV line.

The 161 kV switchyard is connected to one 161 kV line which terminates in a switchyard near Auburn, Nebraska, and the 345/161 kV, 300 MVA auto-transformer which connects to the 345 kV switchyard. The 345 kV switchyard has five lines which terminate in switchyards near Tarkio, Missouri; Hallam, Nebraska; St. Joseph, Missouri; Fairport, Missouri; and Nebraska City, Nebraska. The ESST is fed by a 69 kV line which is part of a subtransmission grid of the Omaha Public Power District. If the normal station service transformer (NSST) (powered by the main generator) is lost, the SSST, which is normally energized, will automatically energize 4160 volt buses 1A and 1B, as well as their connected loads, including critical buses 1 F & 1G. If the SSST fails to energize the critical buses, the ESST, which is normally energized, will automatically energize both critical buses. If the ESST were also to fail, the emergency diesel generators would automatically energize their respective buses. A detailed description of the offsite power network and circuits to the onsite Class 1 E critical buses is found in the USAR, Sections VIII-2.0 and VIII-3.0 (Ref. 2).

Cooper B 3.8-1 02/22/11

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

The 31 day Frequency is adequate to ensure that a sufficient supply of fuel oil is available, since low level alarms are provided and facility operators would be aware of any large uses of fuel oil during this period.

SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day tanks once every 31 days eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation. Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequency is consistent with Regulatory Guide 1.137 (Ref. 11). This SR is for preventive maintenance. The presence of water does not necessarily represent a failure of this SR provided that accumulated water is removed during performance of this Surveillance.

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

The Frequency for this SR corresponds to the testing requirements for pumps as contained in the ASME Code for Operation and Maintenance of Nuclear Power Plants (Ref. 14).

Cooper B 3.8-20 04/28/10

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.1.8 Transfer of each 4.16 kV critical bus power supply from the normal offsite circuit to the alternate offsite circuit demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. The 18 month Frequency of the Surveillance is based on engineering judgment taking into consideration the plant conditions required to perform the Surveillance, and is intended to be consistent with expected fuel cycle lengths. Operating experience has shown that these components usually pass the SR when performed on the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note. The reason for the Note is that, during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, plant safety systems.

Credit may be taken for unplanned events that satisfy this SR.

SR 3.8.1.9 Consistent with IEEE 387-1995 (Ref. 15), Section 7.5.9 and Table 3, this SR requires demonstration once per 18 months that the DGs can start and run continuously at full load capability for an interval of not less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> -6 hours of which is at a load equivalent to 90-100% of the continuous rating of the DG, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of which is at a load equivalent to 105% to 110% of the continuous duty rating of the DG. The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelube and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.

A load band of 90-100% accident load is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. Generator loadings less than 90%

occurring during the first 10 seconds of accident loading are bounded by the test conditions of 90 to 100% load and are well within the generator capability curves.

Cooper B 3.8-21 03125/11

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)

The 18 month Frequency is conservative with respect to the recommendations of IEEE 387-1995 (Ref. 15). IEEE 387-1995 (Ref.

15), Section 7.5.9 and Table 3, require this SR to be performed during refueling outages once per 24 months. The 18 month Frequency takes into consideration plant conditions required to perform the Surveillance; and is intended to be consistent with expected fuel cycle lengths.

This Surveillance has been modified by three Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that would challenge continued steady state operation and, as a result, plant safety systems.

Note 3 ensures that the DG is tested under load conditions that are as close to worst case design basis conditions as possible. When synchronized with offsite power, testing should be performed at a power factor of < 0.89. This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions. Under certain conditions, however, Note 3 allows the surveillance to be conducted at a power factor other than < 0.89. These conditions occur when grid voltage is high, and the additional field excitation needed to obtain a power factor of < 0.89 results in voltages on the emergency busses that are too high. Under these conditions, the power factor should be maintained as close as practicable to 0.89 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.89 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained as close as practicable to 0.89 without exceeding the DG excitation limits. Credit may be taken for unplanned events that satisfy this SR.

SR 3.8.1.10 Under LOCA conditions and loss of offsite power, loads are sequentially connected to the bus by a timed logic sequence. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. The 10%

load sequence time interval tolerance ensures that sufficient time exists for the DG to restore frequency and voltage prior to applying the next Cooper B 3.8-22 03/25/11

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) load and that safety analysis assumptions regarding ESF equipment time delays are not violated. Reference 2 provides a summary of the automatic loading of ESF buses.

The Frequency of 18 months is consistent with the recommendations of Regulatory Guide 1.108 (Ref. 10), paragraph 2.a.(2); takes into consideration plant conditions required to perform the Surveillance; and is intended to be consistent with expected fuel cycle lengths.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR.

SR 3.8.1.11 In the event of a DBA coincident with a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.

This Surveillance demonstrates DG operation during a loss of offsite power actuation test signal in conjunction with an ECCS initiation signal.

This test verifies all actions encountered from the loss of offsite power and loss of coolant accident, including shedding of the nonessential loads and energization of the emergency buses and respective loads from the DG. It further demonstrates the capability of the DG to automatically maintain the required voltage and frequency.

The DG auto-start time of 14 seconds is derived from requirements of the accident analysis for responding to a design basis large break LOCA. The Surveillance should be continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability has been achieved.

The requirement to verify the connection and power supply of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or systems are not capable of being operated at full flow. In lieu of actual demonstration of connection and Cooper B 3.8-23 03/25/11 1

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

The Frequency of 18 months takes into consideration plant conditions required to perform the Surveillance and is intended to be consistent with an expected fuel cycle length of 18 months.

This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil being periodically circulated and temperature maintained consistent with manufacturer recommendations. The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy this SR.

REFERENCES 1. USAR, Section VIII-1.0.

2. USAR, Section VIII-2.0 and VIII-3.0.
3. Safety Guide 9, Revision 0, March 1971.
4. USAR, Chapter VI.
5. USAR, Chapter XIV.
6. 10 CFR 50.36(c)(2)(ii).
7. Generic Letter 84-15.
8. Regulatory Guide 1.93.
9. Regulatory Guide 1.9, Revision 3, July 1993.
10. Regulatory Guide 1.108.
11. Regulatory Guide 1.137.

Cooper B 3.8-24 03/25/11 1

AC Sources - Operating B 3.8.1 BASES REFERENCES (continued)

12. ANSI C84.1, 1970.
13. USAR, Section VIII-5.2.
14. ASME Code for Operation and Maintenance of Nuclear Power Plants.
15. IEEE Standard 387, 1995.

Cooper B 3.8-25 04/28/10

Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES ACTIONS (continued) particulates does not mean failure of the fuel oil to burn properly in the diesel engine, since particulate concentration is unlikely to change significantly between Surveillance Frequency intervals, and since proper engine performance has been recently demonstrated (within 31 days), it is prudent to allow a brief period prior to declaring the DGs inoperable.

The 7 day Completion Time allows for further evaluation, resampling, and re-analysis of the DG fuel oil.

D. 1 With the new fuel oil properties defined in the Bases for SR 3.8.3.3 not within the required limits, a period of 30 days is allowed for restoring the stored fuel oil properties. This period provides sufficient time to test the stored fuel oil to determine that the new fuel oil, when mixed with previously stored fuel oil, remains acceptable, or to restore the stored fuel oil properties. This restoration may involve feed and bleed procedures, filtering, or combination of these procedures. Even if a DG start and load was required during this time interval and the fuel oil properties were outside limits, there is high likelihood that the DG would still be capable of performing its intended function. If the new fuel has not yet been added to the fuel oil storage tanks, entry into this Condition is not necessary.

E. 1 With pressure at least 200 psig in at least one starting air receiver, sufficient capacity for multiple DG start attempts in accordance with References 7 and 9 exists. As long as the pressure is at least 125 psig in at least one starting air receiver, there is capacity for at least one start attempt, and the DG can be considered OPERABLE while the air receiver pressure is restored to the required limit. A period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> is considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable. This period is acceptable based on the remaining air start capacity, the fact that most DG starts are accomplished on the first attempt, and the low probability of an event during this brief period.

F. 1 With a Required Action and associated Completion Time of Condition A, B, C, D, or E not met, or the stored diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than addressed by Conditions A, B, C, D, or E, the associated DG(s) may be incapable of performing its intended function and must be immediately declared inoperable.

Cooper B 3.8-36 08/23/11

Diesel Fuel Oil, Lube Oil, and Starting Air B 3.8.3 BASES REFERENCES (continued)

5. USAR, Chapter XIV.
6. 10 CFR 50.36(c)(2)(ii).
7. USAR, Section VIII-5.3.3.
8. ASTM Standards: D4057-1988E1; D975-1989a; D4176-1991; D1796-1983; D1552-1990; D2622-1992; and D2276-1989.
9. NEDC 11-072, DGSA Accumulator Sizing Basis

(.

Cooper B 3.8-41 08/23/11

DC Sources - Operating B 3.8.4 BASES BACKGROUND (continued) physically and electrically from the other subsystems to ensure that a.

single failure in one subsystem does not cause a failure in a redundant subsystem. There is no sharing between redundant Class I E subsystems.such as batteries, battery chargers, or distribution panels.

The batteries for DC electrical power subsystems are sized to produce required capacity at 90% of nameplate rating. The life of the batteries will be adjusted based on engineering evaluations of each battery's capacity trend data as the batteries age. The minimum design voltage limits are 105 V and 210 V for the 125 V DC and the 250 V DC subsystems, respectively.

Each DC electrical power subsystem battery charger has ample power output capacity for the steady state operation of connected loads required during normal operation, while at the same time maintaining its battery bank fully charged. Each station service battery charger has sufficient capacity to restore the battery from the design minimum charge to its fully charged state within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while supplying normal steady state loads (Ref. 3).

APPLICABLE SAFETY ANALYSES The initial conditions of Design Basis Accident (DBA) and transient analyses in the USAR, Chapter XIV (Ref. 4) assume that Engineered Safety Feature (ESF) systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the DGs, emergency auxiliaries, and control and switching during all MODES of operation. The OPERABILITY of the DC subsystems is consistent with the initial assumptions of the accident analyses and is based upon meeting the design basis of the unit. This includes maintaining DC sources OPERABLE during accident conditions in the event of:

a. An assumed loss of all offsite AC power or all onsite AC power; and
b. A worst case single failure.

The DC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 5).

Cooper B 3.8-43 04/22/10

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS SR 3.8.4.1 Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the charging system and the ability of the batteries to perform their intended function.. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery (or battery cell) and maintain the battery (or battery cell) in a fully charged state. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the initial voltages assumed in the battery sizing calculations. The 7 day Frequency is conservative when compared with the manufacturer's recommendations and IEEE-450 (Ref. 7).

SR 3.8.4.2 Visual inspection to detect corrosion of the battery cells and connections, or measurement of the resistance of each inter-cell, inter-rack, inter-tier, and terminal connection, provides an indication of physical damage or abnormal deterioration that could potentially degrade battery performance.

The limits for battery connection resistance are specified in Table 3.8.4-1.

For inter-cell, inter-tier, and terminal connections, the limits are 150 micro-ohm. For inter-rack connections, the limit is 280 micro-ohm.

The total resistance of the batteries is also monitored. This total resistance is the sum of the inter-cell connectors, the inter-tier cables and connectors, the inter-rack cables and connectors, and the terminal connections. The limits for total resistance in the load and voltage studies are 3355 micro-ohm for the 125 volt batteries (Ref. 11 and 12),

6595 micro-ohm for Division 1 of the 250 volt battery (Ref. 13), and 6775 micro-ohm for Division 2 of the 250 volt battery (Ref. 14). The total resistance limits in Table 3.8.4-1 are conservative two significant digit expressions of the calculated limits.

The Frequency for these inspections, which can detect conditions that can cause power losses due to resistance heating, is 92 days. This Frequency is considered acceptable based on operating experience related to detecting corrosion trends.

Cooper B 3.8-46 04/22/10

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.8.4.3 Visual inspection of the battery cells, cell plates, and battery racks-provides-an indication of physical damage Orabnormal deterioration that could potentially degrade battery performance. The presence of physical damage or deterioration does not necessarily represent a failure of this SR, provided an evaluation determines that the physical damage or deterioration does not affect the Operability of the battery (its ability.to perform its design function). The 18 month Frequency for the Surveillance is based on engineering judgement. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency. Therefore, the Frequency has been concluded to be acceptable from a reliability standpoint.

SR 3.8.4.4 and SR 3.8.4.5 Visual inspection and resistance measurements of inter-cell, inter-rack, inter-tier, and terminal connections provides an indication of physical damage or abnormal deterioration that could indicate degraded battery condition. The anti-corrosion material is used to help ensure good electrical connections and to reduce terminal deterioration. The visual inspection for corrosion is not intended to require removal of and inspection under each terminal connection.

The removal of visible corrosion is a preventive maintenance SR. The presence of visible corrosion does not necessarily represent a failure of this SR, provided visible corrosion is removed during performance of this Surveillance.

The limits for battery connection resistance are specified in Table 3.8.4-1.

For inter-cell, inter-tier, and terminal connections, the limits are 150 micro-ohm. For inter-rack connections, the limit is 280 micro-ohm.

The total resistance of the batteries is also monitored. This total resistance is the sum of the inter-cell connectors, the inter-tier cables and connectors, the inter-rack cables and connectors, and the terminal connections. The limits for total resistance in the load and voltage studies are 3355 micro-ohm for the 125 volt batteries (Ref. 11 and 12),

6595 micro-ohm for Division 1 of the 250 volt battery (Ref. 13), and 6775 micro-ohm for Division 2 of the 250 volt battery (Ref. 14). The total resistance limits in Table 3.8.4-1 are conservative two significant digit expressions of the calculated limits.

Cooper B 3.8-47 04/22/10

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)

The 18 month Frequency for the Surveillances is based on engineering judgment. Operating experience has shown that.these components usually pass the SR when performed at the 18 month Frequency.

Therefore, the Frequency has been concluded to be acceptable from a reliability standpoint.

SR 3.8.4.6 Battery charger capability requirements are based on the design capacity of the chargers (Ref. 3). According to Regulatory Guide 1.32 (Ref. 8),

the battery charger supply is required to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied.

The Frequency is acceptable, given the unit conditions required to perform the test and the other administrative controls existing to ensure adequate charger performance during these 18 month intervals. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths.

S R 3.8.4.7 A battery service test is a special test of the battery's capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length corresponds to the design duty cycle requirements as specified in design calculations.

The Frequency of 18 months is consistent with the recommendations of Regulatory Guide 1.32 (Ref. 8) and Regulatory Guide 1.129 (Ref. 9),

which state that the battery service test should be performed during refueling operations or at some other outage, with intervals between tests not to exceed 18 months.

This SR is modified by two Notes. .Note 1 allows the performance of a modified performance discharge test in lieu of a service test once per 60 months. The substitution is acceptable because a modified performance discharge test represents a more severe test of battery capacity than SR 3.8.4.7.

Cooper B 3.8-48 04/22/10

DC Sources - Operating B 3.8.4 BASES.

SURVEILLANCE REQUIREMENTS (continued)

The reason for Note 2 is that performing the Surveillance would remove a required DC electrical, power subsystem from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that. satisfy the Surveillance.-

SR 3.8.4.8 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

A battery modified performance discharge test is a simulated duty cycle consisting of just two rates; the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance discharge test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a rated one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test should remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test. Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.4.8; however, only the modified performance discharge test may be used to satisfy SR 3.8.4.8 while satisfying the requirements of SR 3.8.4.7 at the same time.

The acceptance criteria of > 90% capacity for this Surveillance is conservative with respect to IEEE-450 (Ref. 7) and IEEE-485 (Ref. 10).

These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.

Cooper B 3.8-49 04/22/10 1

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued)

The Frequency for-this test is normally 60 months. If the battery shows degradation, or. if the battery has reached 15 years (85% of its expected life) and capacity is < 100% of.the manufacturer's rating, the Surveillance Frequency is reduced to 18 months... However, if the battery shows no degradation but has reached 85% of its.expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity

> 100% of the manufacturer's rating. Degradation is indicated, according to IEEE-450 (Ref. 7), When the battery capacity drops by more than 10%

relative to its capacity on the previous performance tests or when it is below 90% of the manufacturer's rating. However, at Cooper Nuclear Station degradation is defined when the battery capacity drops by more than 5% relative to the capacity on the previous performance test or when the battery capacity < 95% of the manufacturer's rating. This more restrictive definition of degradation is necessary to ensure that the decision can be made for battery replacement before the > 90% capacity technical specification is violated. The 60 month frequency is consistent with the recommendations in IEEE-450 (Ref. 7). The 18 month and 24 month Frequencies are derived from the recommendations in IEEE-450 (Ref. 7)

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DC electrical power subsystem from service, perturb the electrical distribution system, and challenge safety systems. Credit may be taken for unplanned events that satisfy the Surveillance.

REFERENCES 1. USAR, Section VIII-6.2.

2. Regulatory Guide 1.6.
3. IEEE Standard 308, 1970.
4. USAR, Chapter XIV.
5. 10 CFR 50.36(c)(2)(ii).
6. Regulatory Guide 1.93.
7. IEEE Standard 450, 1995.
8. Regulatory Guide 1.32, February 1977.

Cooper B 3.8-50 04/22/10

DC Sources - Operating B 3.8.4 BASES REFERENCES (continued)

9. Regulatory, Guide 1.129, December 1974.
10. IEEE Standard 485, .1983.
11. NEDC 87-1 31 C, 125 VDC Division I Load and Voltage Study.
12. NEDC 87-131 D, 125 VDC Division I1Load and Voltage Study. * .1
13. NEDC 87-131A, 250 VDC Division I Load and Voltage Study.
14. NEDC 87-131B, 250 VDC Division I1Load and Voltage Study.

Cooper B 3.8-51 04/22/10

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