ML102150035

Technical Specifications Bases Revision 52 Update

ML102150035
Person / Time
Site:	Palo Verde
Issue date:	07/21/2010
From:	Weber T Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-06229-TNW/RAS/CJS
Download: ML102150035 (87)

Text

{{#Wiki_filter:Technical Specification 5.5.14 Palo Verde Nuclear Generating Station Thomas N. Weber Department Leader Regulatory Affairs Tel. 623-393-5764 Fax 623-393-5442 Mail Station 7636 PO Box 52034 Phoenix, Arizona 85072-2034 ID#: 102-06229-TNW/RAS/CJS July 21, 2010 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 Docket Nos. STN 50-528/529/530 Technical Specifications Bases Revision 52 Update Pursuant to PVNGS Technical Specification (TS) 5.5.14, "Technical Specifications Bases Control Program," Arizona Public Service Company (APS) is submitting changes to the TS Bases incorporated into Revision 52, implemented on July 16, 2010. The revision insertion instructions and replacement pages are provided in the enclosure. No commitments are being made to the NRC by this letter. Should you need further information regarding this submittal, please contact Russell A. Stroud, Licensing Section Leader, at (623) 393-5111. Sincerely, ýJ, 6Y M-,, TNW/RAS/CJS/gat

Enclosure:

PVNGS Technical Specification Bases Revision 52 Insertion Instructions and Replacement Pages cc: E. E. Collins Jr. J. R. Hall L. K. Gibson R. I. Treadway NRC Region IV Regional Administrator (enclosure) NRC NRR Senior Project Manager (enclosure) NRC NRR Project Manager (enclosure) NRC Senior Resident Inspector for PVNGS (enclosure) /h( /Ou-A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway 0 Comanche Peak 0 Diablo Canyon 0 Palo Verde 0 San Onofre

• South Texas 0 Wolf Creek

ENCLOSURE PVNGS Technical Specification Bases Revision 52 Insertion Instructions and Replacement Pages

Digitally signed by Stephenson, Carl J(Z05778) Stephenson, JDN: cn=Stephenson, Carl J(Z05778) Reason This is an accurate copy of the original ,,.document. Carl J(Z05778) -Date: 010.07.13 14:47:30 -0700' Insertion Instructions for the Technical Specifications Bases Revision 52 REMOVE PAGES Cover page List of Effective Pages 1/2 through 7/8 B 3.1.5-1 / B 3.1.5-2 through B 3.1.5-11 / Blank B 3.1.8-1 / B 3.1.8-2 through B 3.1.8-5 / Blank B 3.2.2-1 / B 3.2.2-2 B 3.2.2-3 / B 3.2.2-4 B 3.2.3-1 / B 3.2.3-2 B 3.2.3-3 / B 3.2.3-4 B 3.2.4-1 / B 3.2.4-2 B 3.2.4-3 / B 3.2.4-4 B 3.2.5-1 / B 3.2.5-2 B 3.2.5-3 / B 3.2.5-4 B 3.2.5-5 / B 3.2.5-6 B 3.3.1-21 / B 3.3.1-22 B 3.3.1-23 / B 3.3.1-24 B 3.3.5-23 / B 3.3.5-24 B 3.4.3-1 / B 3.4.3-2 through B 3.4.3-7 / B 3.4.3-8 B 3.4.6-3 / B 3.4.6-4 B 3.4.6-5 / Blank B 3.4.7-3 / B 3.4.7-4 B 3.4.7-7 / Blank INSERT PAGES Cover page List of Effective Pages 1/2 through 7/8 B 3.1.5-1 / B 3.1.5-2 through B 3.1.5-11 / B 3.1.5-12 B 3.1.8-1 / B 3.1.8-2 through B 3.1.8-5 / Blank B 3.2.2-1 / B 3.2.2-2 B 3.2.2-3 / B 3.2.2-4 B 3.2.3-1 / B 3.2.3-2 B 3.2.3-3 / B 3.2.3-4 B 3.2.4-1 / B 3.2.4-2 B 3.2.4-3 / B 3.2.4-4 B 3.2.5-1 / B 3.2.5-2 B 3.2.5-3 / B 3.2.5-4 B 3.2.5-5 / B 3.2.5-6 B 3.3.1-21 / B 3.3.1-22 B 3.3.1-23 i B 3.3.1-24 B 3.3.5-23 / B 3.3.5-24 B 3.4.3-1 / B 3.4.3-2 through B 3.4.3-7 / B 3.4.3-8 B 3.4.6-3 / B 3.4.6-4 B 3.4.6-5 / Blank B 3.4.7-3 / B 3.4.7-4 B 3.4.7-7 / Blank I

B 3.4.11-3 / B 3.4.11-4 B 3.4.13-3 / B 3.4.13-4 through B 3.4.13-7 / B 3.4.13-8 B 3.4.13-11 / Blank B 3.7.12-3 / B 3.7.12-4 B 3.7.17-1 / B 3.7.17-2 B 3.7.17-5 / B 3.7.17-6 B 3.4.11-3./ B 3.4.11-4 B 3.4.13-3 / B 3.4.13-4 through B 3.4.13-7 / B 3.4.13-8 B 3.4.13-11 / Blank B 3.7.12-3 / B 3.7.12-4 B 3.7.17-1 / B 3.7.17-2 B 3.7.17-5 / B 3.7.17-6 2

PVNGS Palo Verde Nuclear Generating Station Units 1, 2, and 3 Technical Specification Bases Revision 52 July 16, 2010 Oigitellyj signed by Stephenson, Carl Stephenson, ,ON 5 8 Carl (Z05 Reason: Th~s is an accurate copy of Ithe Carl:Jc(Z-5778)sorig d crument c'tlJ(0Z0578* 1..°"Io 1*2 7 -0700o Date: 2010.07.13 14:20:37 -0OM0

TECHNICAL SPECIFICATION BASES LIST OF EFFECTIVE PAGES Page Rev. Page Rev NO. NO. No. No. B 2.1.1-1 0 B 3.1.4-2 31 B 2.1.1-2 0 B 3.1.4-3 0 B 2.1.1-3 37 B 3.1.4-4 0 B 2.1.1-4 21 B 3.1.4-5 0 B 2.1.1-5 23 B 3.1.5-1 0 B 2.1.2-1 0 B 3.1.5-2 52 B 2.1.2-2 31 B 3.1.5-3 52 B 2.1.2-3 0 B 3.1.5-4 52 B 2.1.2-4 23 B 3.1.5-5 52 B 2.1.2-5 0 B 3.1.5-6 52 B 3.0-1 49 B 3.1.5-7 52 B 3.0-2 0 B 3.1.5-8 52 B 3.0-3 0 B 3.1.5-9 52 B 3.0-4 0 B 3.1.5-10 52 B 3.0-5 42 B 3.1.5-11 52 B 3.0-6 48 B 3.1.5-12 52 B 3.0-7 48 B 3.1.6-1 0 B 3.0-8 42 B 3.1.6-2 46 B 3.0-9 42 B 3.1.6-3 42 B 3.0-10 42 B 3.1.6-4 42 B 3.0-11 42 B 3.1.6-5 46 B 3.0-12 42 B 3.1.6-6 46 B 3.0-13 42 B 3.1.7-1 0 B 3.0-14 49 B 3.1.7-2 0 B 3.0-15 50 B 3.1.7-3 50 B 3.0-16 50 B 3.1.7-4 48 B 3.0-17 50 B 3.1.7-5 25 B 3.0-18 49 B 3.1.7-6 0 B 3.0-19 49 B 3.1.7-7 0 B 3.0-20 49 B 3.1.7-8 51 B 3.0-21 49 B 3.1.7-9 0 B 3.0-22 49 B 3.1.8-1 52 B 3.1.1-1 28 B 3.1.8-2 52 B 3.1.1-2 0 B 3.1.8-3 52 B 3.1.1-3 43 B 3.1.8-4 52 B 3.1.1-4 43 B 3.1.8-5 52 B 3.1.1-5 27 B 3.1.9-1 0 B 3.1.1-6 31 B 3.1.9-2 0 B 3.1.2-1 28 B 3.1.9-3 0 B 3.1.2-2 0 B 3.1.9-4 0 B 3.1.2-3 43 B 3.1.9-5 47 B 3.1.2-4 28 B 3.1.9-6 1 B 3.1.2-5 0 B 3.1.10-1 0 B 3.1.2-6 43 B 3.1.10-2 28 B 3.1.2-7 12 B 3.1.10-3 0 B 3.1.2-8 47 B 3.1.10-4 37 B 3.1.2-9 0 B 3.1.10-5 37 B 3.1.3-1 0 B 3.1.10-6 0 B 3.1.3-2 0 B 3.1.11-1 0 B 3.1.3-3 0 B 3.1.11-2 28 B 3.1.3-4 0 B 3.1.11-3 0 B 3.1.3-5 0 B 3.1.11-4 34 B 3.1.3-6 0 B 3.1.11-5 0 B 3.1.4-1 0 B 3.2.1-1 50 PALO VERDE UNITS 1, 2, AND 3 1Revision 52 July 16, 2010

TECHNICAL SPECIFICATION BASES LIST OF EFFECTIVE PAGES Page No. Page No. Rev. No. Rev No. No. 3.2.1-2 3.2.1-3 3.2.1-4 3.2.1-5 3.2.1-6 3.2.1-7 3.2.1-8 3.2.2-1 3.2.2-2 3.2.2-3 3.2.2-4 3.2.2-5 3.2.2-6 3.2.2-7 3.2.3-1 3.2.3-2 3.2.3-3 3.2.3-4 3.2.3-5 3.2.3-6 3.2.3-7 3.2.3-8 3.2.3-9 3.2.3-10 3.2.4-1 3.2.4-2 3.2.4-3 3.2.4-4 3.2.4-5 3.2.4-6 3.2.4-7 3.2.4-8 3.2.4-9 3.2.4-10 3.2.5-1 3.2.5-2 3.2.5-3 3.2.5-4 3.2.5-5 3.2.5-6 3.2.5-7 3.3.1-1 3.3.1-2 3.3.1-3 3.331-4 3.3.1-5 3.3.1-6 3.3.1-7 3.3.1-8 3.3.1-9 3.3.1-10 3.3.1-11 3.3.1-12 3.3.1-13 10 28 0 0 0 0 0 52 10 0 52 1 0 0 52 10 0 52 0 0 0 0 0 0 52 10 0 52 25 25 27 48 48 31 52 10 0 52 0 52 0 35 25 25 25 A 25 27 25 25 34 35 35 35 35 3.3.1-14 3.3.1-15 3.3.1-16 3.3.1-17 3.3.1-18 3.3.1-19 3.3.1-20 3.3.1-21 3.3.1-22 3.3.1-23 3.3.1-24 3.3.1-25 3.3.1-26 3.3.1-27 3.3.1-28 3.3.1-29 3.3.1-30 3.3.1-31 3.3. 1-32 3.3.1-33 3.3.1-34 3.3.1-35 3.3.1-36 3.3.1-37 3.3.1-38 3 3.1-39 3.3.1-40 3.3.1-41 3.3.1-42 3.3. 1-43 3.3.1-44 3.3.1-45 3.3.1-46 3.3.1-47 3.3.1-48 3.3.1-49 3.3.1-50 3.3.1-51 3.3.1-52 3.3.1-53 3.3.1-54 3.3.1-55 3.3.1-56 3.3.1-57 3.3.1-58 3.3.1-59 3.3.1-60 3.3.2-1 3.3.2-2 3.3.2-3 3.3.2-4 3.3.2-5 3.3.2-6 3.3.2-7 35 35 35 35 35 35 35 35 52 52 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 35 38

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TECHNICAL SPECIFICATION BASES LIST. -OO EFFECTIVE PAGES Page No. Rev. No.- B 3.3.2-8 B 3.3.2-9 B 3.3.2-10 B 3.3.2-11 B 3.3.2-12 B 3.3.2-13 B 3.3.2-14 B 3.3.2-15 B 3.3.2-16 B 3.3.2-17 B 3.3.2-18 B 3.3.3-1 B 3.3.3-2 B 3.3.3-3 B,3-.3.3-4 B 3.3.3-5 B 3.3,.3-6 B,31 .33-7 B 3 '.3 .3-8 B 3.3.3-9 B 3.3.3-10 B 3.3.3-11 B 3.3.3-12 B 3.3.3-13 B 3.3.3-14 B 3.3.3-15 B 3.3.3-16 B 3.3.3-17 B 3.3.3-18 B 3.3.3-19 B 3.3.3-20 B 3.3.3-21 B 3.3.4-1 B 3.3.4-2 B 3.3.4-3 B 3.3.4-4 B 3.3.4-5 B 3.3.4-6 B 3.3.4-7 B 3.3.4-8 B 3.3.4-9 B 3.3.4-10 B 3.3.4-11 B 3.3.4-12 B 3.3.4-13 B 3.3.4-14 B 3.3.4-15 B 3.3.5-1 B 3.3.5-2 B 3.3.5-3 B 3.3.5-4 B 3.3.5-5 B 3.3.5-6 B 3.3.5-7 35 50 38 42 42 51 51 35 35 35 35 25 27 25 25 25 25 27 27 27 46 25 25 25 46 27 51 51 51 51 51 27 0 0 0 0 0 31 0 0 0 0 0 0 0 0 0 0 0 0

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TECHNICAL SPECIFICATION BASES LIST OF EFFECTIVE PAGES Page Rev. Page No. No. No... Rev No. 3.3.8-1 3.3.8-2 3.3.8-3 3.3.8-4 3.3.8-5 3.3.8-6 3.3.8-7 3.3.8-8 3.3.9-1 3.3.9-2 3.3.9-3 3.3.9-4 3.3.9-5 3.3.9-6 3.3.9-7 3.3.10-1 3.3.10-2 3.3.10-3 3.3.10-4 3.3.10-5 3.3.10-6 3.3.10-7 3.3.10-8 3.3.10-9 3.3.10-10 3.3.10-11 3.3. 10-12 3.3.10-13 3.3.10-14 3.3. 10-15 3.3.10-16 3.3.10-17 3.3.10-18 3.3.10-19--ý

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3.3.12-5 3A3.12-6' 3.4.1-1 3.4.1-2 3.4.1-3 3.4.1-4 0 44 0 0 0 51 0 44 48 48 21 10 51 0 0 0 0 0 0 18 0 0 14 14 51 50 50 50 50 50 50 50 50 51 50 50 32 0 2 2 42 42 51 50 15 50 37 37 51 6: 10 28 0 0 3.4.1-5 3.4.2-1 3.4.2-2 3.4.3-1 3.4.3-2 3.4.3-3 3.4.3-4 3.4.3-5 3.4.3-6 3.4.3-7 3.4.3-8 3.4.4-1 3.4.4-2 3.4.4-3 3.4.4-4 3.4.5-1 3.4.5-2 3.4.5-3 3.4.5.-4 3.4.5-5 3.4.6-1 3.4.6-2 3.4.6-3 3.4.6-4 3.4.6-5 3.4.7-1 3.4.7-2 3.4.7-3 3.4.7-4 3.4.7-5 3.4.7-6 3.4.7-7 3.4.8-1 3.4.8-2 3.4.8-3 3.4.9-1 3.4.9-2 3.4.9-3 '4 3.4.9-4 3.4.9-5 3.4.9-61 3.4.10-3.4.10-2 3.4.10-3 3.4.10-4 3.4.11-1 3.4.11-3.4.11-3 3.4.11-45 3.4.11-5-, 3.4.11-6 3.4.12-1 3.4.12-23: 3.4.12-3* 0 7 1 52 52 0 52 .52 0 52 52 0 50 7 0 0 38 38 0 6 0 6 52 6 52 0 6 52 38 00 52 0 6 6 41 31 41 41 0 0 50 7 0 0 0 7 0 52 0 0 1 34 48 C PALO VERDE UNITS 1, 2, AND 3 .4 Revision 52 July 16, 2010

TECHNICAL SPECIFICATION BASES LIST OF EFFECTIVE PAGES Page No. Page No. Rev. No. Rev No. B 3.4.12-4 B 3.4.12-5 B 3.4.13-1 B 3.4.13-2 B 3.4.13-3 B 3.4.13-4 B 3.4.13-5 B 3.4.13-6 B 3.4.13-7 B 3.4.13-8 B 3.4.13-9 B 3.4.13-10 B 3.4.13-11 B 3.4.14-1 B 3.4.14-2 B 3.4.14-3 B 3.4.14-4 B 3.4.14-5 B 3.4.14-6 B 3.4.14-7 B.3.4.14-8 B 3.4.15-1 B 3.4.15-2 B 3.4.15-3 B 3.4.15-4 B 3.4.15-5 B 3.4.15-6 B 3.4.15-7 B 3.4.16-1 B 3.4.16-2 B 3.4.16-3 B 3.4.16-4 B 3.4.16-5 B 3.4.16-6 B 3.4.17-1 B 3.4.17-2 B 3.4.17-3 B 3.4.17-4 B 3.4.17-5 B 3.4.17-6 B 3.4.18-1 B 3.4.18-2 B 3.4.18-3 B 3.4.18-4 B 3.4.18-5 B 3.4.18-6 B 3.4.18-7 B 3.4.18-8 B 3.5.1-1 B 3.5.1-2 B 3.5.1-3 B 3.5.1-4 B 3.5.1-5 B 3.5.1-6 0 31 0 0 1 52 52 0 52 52 42 42 52 0 34 34 38 38 38 38 38 0 48 0 0 0 35 35 2 10 0 42 0 0 0 27 42 42 0 0 38 40 38 38-38. 38* 38 38 0, 48" 7 0c 00 B 3.5.1-7 B 3.5.1-8 B 3.5.1-9 B 3.5.1-10 B 3.5.2-1 B 3.5.2-2 B 3.5.2-3 B 3.5.2-4 B 3.5.2-5 B 3.5.2-6 B 3.5.2-7 B 3.5.2-8 B 3.5.2-9 B 3.5.2-10 B 3.5.3-1 B 3.5.3-2 B 3.5.3-3 B 3.5.3-4 B 3.5.3-5 B 3.5.3-6 B 3.5.3-7 B 3.5.3-8 B 3.5.3-9 B 3.5.3-10 B 3.5.4-1 B 3.5.4-2 B 3.5.4-3 B 3.5.5-1 B 3.5.5-2 B 3.5.5-3 B 3,5.5-4 B 3.5.5-5 B 3.5.5-6 B 3.5.5-7 B 3.5.5-8 B 3.5.5-9 B 3.5.5-10" B 3.5.5-11 B 3.5.5-12 B 3.5.5-13 B 3.5.5-14 B 3.5.5-15 B 3.5.5-16 B 3.5.6-1 B 3.5.6-2 B 3.5.6-3 B 3.5.6-4 B 3.5.6-5 B 3.6.1-1 B 3.6.1-2 B 3.6.1-3 B 3.6.1-4 B 3.6.1-5 B 3.6.2-1 11 035 035 0 0 0 22 1 35 0 48 0 0 2 2 1 0 2 15 42 51 51 51 51-51.' 51*" 51~ 51' 48'. 4& 48 48 48 48) 48

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TECHNICAL SPECIFICATION BASES

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No. Page No. Rev. No. B 3.6.2-2 B 3.6.2-3 B 3.6.2-4 B 3.6.2-5 B 3.6.2-6 B 3.6.2-7 B 3.6.2-8 B 3.6.3-1 B 3.6.3-2 B 3.6.3-3 B 3.6.3-4 B 3.6.3-5 B 3.6.3-6 B 3.6.3-7 B 3.6.3-8 B 3.6.3-9 B 3.6.3-10 B 3.6.3-11 B 3.6.3-12 B 3.6.3-13 B 3.6.3-14 B 3.6.3-15 B 3.6.3-16 B 3.6.3-17 B 3.6.3-18 B 3.6.3-19 B 3.6.4-1 B 3.6.4-2 B 3.6.4-3 B 3.6.5-1 B 3.6.5-2 B 3.6.5-3 B 3.6.5-4 B 3.6.6-1 B 3.6.6-2 B 3.6.6-3 B 3.6.6-4 B 3.6.6-5 B 3.6.6-6 B 3.6.6-7 B 3.6.6-8 B 3.6.6-9 B 3.7.1-1 B 3.7.1-2 B 3.7.1-3 B 3.7.1-4 B 3.7.1-5 B 3.7.1-6 B 3.7.2-1 B 3.7.2-2 B 3.7.2-3 B 3.7.2-4 B 3.7.2-5 B 3.7.2-6 49 0 0 0 0 0 0 36 43 49 43 43 43 43 43 43 43 43 43 43 43 43 43 27 43 43 35 38 0 48 0 0 0 38 7 1 01 48 0 50 50 34 34 34 28 40 42 40 40 40 40 2 B 3.7.2-7 B 3.7.2-8 B 3.7.2-9 B 3.7.3-1 B 3.7.3-2 B 3.7.3-3 B 3.7.3-4 B 3.7.3-5 B 3.7.4-1 B 3.7.4-2 B 3.7.4-3 B 3.7.4-4 B 3.7.4-5 B 3.7.5-1 B 3.7.5-2 B 3.7.5-3 B 3.7.5-4 B 3.7.5-5 B 3.7.5-6 B 3.7.5-7 B 3.7.5-8 B 3.7.5-9 B 3.7.5-10 .B 3.7.5-11 B 3.7.6-1 B 3.7.6-2 B 3.7.6-3 B 3.7.6-4 B 3.7.7-1 B 3.7.7-2 B 3.7.7-3 B 3.7.7-4 B 3.7.7-5 B 3.7.8-1 B 3.7.8-2 B 3.7.8-3 B 3.7.8-4 B 3.7.9-1 B 3.7.9-2 B 3.7.9-3 B 3.7.10-1 B 3.7.10-2 B 3.7.10-3 B 3.7.10-4 B 3.7.11-1 B 3.7.11-2 B 3.7.11-3 B 3.7.11-4 B 3.7.11-5 B 3.7.11-6 B 3.7.11-7 .'B 3 -.7 -11'8 B 3.7.11-9 B 3.7.12-1 40 40 40 1 1 37 0 0 50 50 50 50 50 0 0 40 27 42 42 9 9 9 9 9 0 28 28 0 0 1 ,/ 1 1 N: 1 %" 1 1 50 50 51 50 50 50 50 50 1 Lk Cori~ec ted PALO VERDE UNITS 1, 2, AND 3 6 Revision 52 July 16, 2010

TECHNICAL. SPECIFICATION BASES LIST' OFEFFECTIVE PAGES Rev. Page No. No. Page No. Rev No. B 3.7.12-2 B 3.7.12-3 B 3.7.12-4 B 3.7.13-1 B 3.7.13-2 B 3.7.13-3 B 3.7.13-4 B 3.7.13-5 B 3.7.14-1 B 3.7.14-2 B 3.7.14-3 B 3.7.15-1 B 3.7.15-2 B 3.7.16-1 B 3.7.16-2 B 3.7.16-3 B 3.7.16-4 B 3.7.17-1 B 3.7.17-2 B 3.7.17-3 B 3.7.17-4 B 3.7.17-5 B 3.7.17-6 B 3.8.1-1 B 3.8.1i-2 B 3.8.1-3 B 3.8.1-4 B 3.8.1-5 B 3.8.1-6 B 3.8.1-7 B 3.8.1-8 B 3.8.1-9 B 3.8.1-10 B 3.8.1-11 B 3,8.1-12 B 3.8.1-13 B 3.8.1-14 B 3 8.1-15 B 3 8.1-16 B 3 8.1-17 B 3 8.1-18 B 3 8.1-19 B 3 8.1-20 B 3 8.1-21 B 3.8.1-22 B 3.8.1-23 B 3.8.1-24 B 3.8.1-25 B 3.8.1-26 B 3.8.1-27 B 3.8.1-28 B 3.8.1-29 B 3.8.1-30 B 3.8.1-31 21 52 10 0 0 0 0 0 0 21 21 3 3 7 0 0 0 52 3 3 3 3 52 35 2 34 34 20 27 42 50 42 43 50 48 48 48 .48 41 41 41 41 41 41 41. 50' 50 ' 50 50 50 41 41 50' 50 B 3.8.1-32 B 3.8.1-33 B 3.8.1-34 B 3.8.1-35 B 3.8.1-36 B 3.8.1-37 B 3. 8.1-38 B 3.8.1-39 B 3.8.1-40 B 3.8.1-41 B 3.8.1-42 B 3.8.1-43 B 3.8.1-44 B 3.8.1-45 B.3.8.1-46 B.3.8. 1-47 B. 3.8. 1-48 B 3.8.2-1. B 3.8.2-2 B 3.8.2-3 B 3.8.2-4 B 3.8.2-5 B 3.8.2-6 B 3.8.3-1 B 3.8.3-2 B 3.8.3-3 B 3.8.3-4 B 3.8.3-5 B 3.8.3-6 B 3.8.3-7 B 3.8.3-8 B 3.8.3-9 B 3.8.3-10 B 3.8.4-1 B 3.8.4-2 B 3.8.4-3 B 3.8.4-4 B 3.8.4-5 B 3.8.4-6 B 3.8.4-7 B 3.8.4-8 B 3.8.4-9 B 3.8.4-10 B 3.8.4-11 B 3.8.5-1 B 3.8.5-2 B 3.8.5-3 B,3.8-5,-4. B 3.8.5-5 B 3.8.5-6 B 3.8.6-1 B 3.8.6-2 B 3.8.6-3 B 3.8.6-4 45 48 45 .50 50 45 45 45 48 50 45 45 45 45 50 45 45 0 0 0 21 21 0 0 0 50 0 51 51 41 41 41 48 0 37 0 2 22 35 35 35 37 48 1 1 21 21 2 2 0 .0 0 6 Corrected 2, AND 3

PALO VERDE UNITS 1, 7

Revision 52 July 16, 2010

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CEA Alignment B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS: B 3.'1.5 Control Element Assembly (CEA) Alignment BASES BACKGROUND The OPERABILITY (e.g., trippability) of the shutdown and regulating CEAs is an initial assumption in all safety analyses that assume CEA insertion upon reactor trip. Maximum CEA misalignment is an initial assumption.in the safety analyses that directly affects core power distributions and assumptions of available SDM. The applicable criteria for these reactivity, and power distribution design requirements are 10 CFR 50, Appendix Aý, GDC 10 and GDC 26 (Ref. 1) and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Plants" (Ref. 2). Mechanical or electrical failures may.cause a CEA to become inoperable or to become-misaligned from its group: CEA: inoperability or misalignment may cau:se increased,power peaking, due to the asymmetric reactivity distri,,bution'and a reduction in the total, available CEA worth for reactor shutdown. Therefore, CEA alignment and operability.a.re., related to core operation in design power peaking,,limits and -the core design requirement of a minimum SDM. If a CEA(s,) is discovered to be immovable but remains trippable and c aligned, the. CEA is considered to be OPERABLE. At anytime, if a CEA(s) is immovable, a determination of the' trippability (OPERABILITY) of that CEA(s) must be imade, and appropriate action taken. Limits on CEA alignment and operability have beehn established, and all CEA positions are monitored and,, controlled during power operation to, ensure that:.the-power distribution and reactivity limits defined by the design power peaking and SDM limits are pre served. CEAs are moved by their control element drive mechanisms (CEDMs). Each CEDM moves its CEA one'step (approximately 34 inch) at a time, but at varying rates (steps per, minute) depending on the signal output from the Control.ETement Drive Mechanism Control System (CEDMCS). (continued) -PALO VERDE UNITS 1,2,3 .B 3.1-;5-1 REVISION 0

CEA Alignment B 3.1.5 BASES BACKGROUND (conti~nued) !The CEAs are arranged into groups thatt'are radially' symmetric. Therefore, movement of the CEAs does not,.' introduce radial asymmetries in the core power distribution. The shutdownland regulating CEAs provide the required reactivity worth'for immediate reactor shutdown upon a reactor trip. The"regulating CEAs also provide reactivity (power level)-control durihg normal-operation and transients. Their'movement may be automatically controlled by the Reactor Regulating System. Part strength CEAs are notcredited in the safety analyses for:shutting down the reactor,. as are the regulating and shutdown groups. The 0art'strength CEAs are used solely for ASI control. I The axial position of shutdown and regulating CEAs is indicated by two separate and independent systems, which are the Pulse Counting CEA Position Indication System (described in Ref. 4) and the Reed SwitchiCEA Position Indication System (described in Ref. 5). The Pulse Counting CEA Position Indicating System indicates CEA ositi~on to the.actualstep., if each CEA moves One step for each command signal,. However, if each CEA does not fol16wthe commands,';the.sys~tem will incorrectly reflect the position of the affected CEA(s)*.! IThis condition may affect the, operability of COLSS (refer. to.Section 3.2, Power Distribution Limi'ts fort'the -applicable actions) and should be detected; bY; the ý Reed Swi tch Pos-i ti on"I ndi cati on System -thrdughvsurveillanceo 4r alarm. ',',,Ai.though the Reed Switch Position Indi.'cation System is.less pprecise than the Pulse Count*ng CEA Position Indicating System, it is not subject 'to the same error mnecharnismsý. (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-2 REýISION' 52

CEA Alignment B 3.1.5 BASES (continued) APPLICABLE CEA misalignment accidents are analyzed in the safety SAFETY ANALYSES analysis (Ref. 3). The accident analysis defines CEA misoperation as any event, with the exception of sequential group withdrawals, which could result from a single malfunction in,ýthe reactivity control systems. For example, CEA.misalignment may be caused by a malfunction of the CEDM, CEDMCS,- or by, operator error. A stuck'CEA may be caused by mechanical jamming of the CEAfingers..or of the gripper.

• 'Inadvertent withdrawal of a single CEA may be caused by an electrical failurejin the CEA~coil power programmers.

A-dropped CEA could be ca.used.by, an:opening of the electrical circuit of the CEDM holding coil for a full strength, or part.-strength CEA... The acceptance criteria for addressing :CEA inoperability or misalignment are. that: There shall be no violations of: 1> specified aceptab~le fuel design limits, or 2:.i Reactor Coolant System (RCS),pressure boundary 7 integrit.~. ..;,Toensure that these acceptance criteria are met, the CEAs "'shall be.. capable ofI inserting;.the requi red negati ve , reactivity and in ýthe ti1 me peri od !asumed in the accident analysis upon a.,reactor tip. Three types of mi~salgnmeht are distingbished. They are misalignment within deadband (< 6.6 inches), misalignment in excess of deadband, and CEA/subgroup drop. During movement of a group, one CEA may stop moving while the other CEAs in the group continue. This condition may cause excessive power peaking. This misalignment can be within or exceed the deadband. The last type of misalignment occurs when one CEA or subgroup drops partially or fully into the reactor core. This event causes an initial power reduction followed by a return towards the original power due to positive reactivity feedback from the negative moderator temperature coefficient. Increased peaking during the power increase may result in erosion of DNB margin. (continued) ,PALO VERDE UNITS 1,2,3 B 3.1.5-3 REVISION 52

CEA Alignment B 3.1.5 BASES APPLICABLE SAFETY ANALY1 (continued Misalignments within deadband are evaluated to ensure SES specifiedacceptable fbeT design limits (SAFDLs) are notj. exceeded.' Misalignments in excess bf deadband considers"the case of a,singleCEA withdrawn approximately 10 inches from a bank inserted, to its insertionilimit.. Satisfying limits on departure from. nucleate boilin'g *ratjio (DNBR) bounds the situation when a CEA is misaligned frbim.its group by 6.6 inches. The effect of any misoperated.CEA on the core power distribution will be assessed.by the CEA calcul ators, and an appropriately augmented powerdistribution penalty factor will be supplied as input:tothe core protection calculators (CPCs). As the reactor core responds, to-the reactivity changes causedby the misoperated CEA.,and the ensuing reactor coolant and Doppler feedback, effects, the CPCs will initiate a low DNBR or high local power density trip signal if SAFDLs are. approached. Qw, denitytri sina The acci~dent-analysis analvze-a single.four finger full and part strength CEA drp-, *a twel.vef, i'ng'eprdrop, and a subgroup drop. The ýtwelve flinge'r.a6d subgrbup' drops, cau`se larger distortions than,.the four finger drops. !,With CEACS In Service (IS), the subgroup and..twelvefjihger rod drops will result ina penalty fac-tor such th'at 'a CPC trip will occur if SAFDLs are approached. The four fji, geIr CEA drop is protected by the thermal margin-reserved~in COLSS or CPC DNBR*,limit lines (COLR-fi~gures 3.2.4-2 for CEACs IS and 3.2.4-3 for CEACs OOS) whe,,COLSS,is.Out.of Service (OOS). With CEACs OOS, CPCs will not penalizle DNBR nor LPD when CEAs are misaligned; therefore, additional thermal margin is required to be-preserved due to-the-larger -radia.l -power-- distortion"associated with twelve finger and subgroup drops. The most rapid approach to the DNBR. SAFDL or the fuel centerli ie i'elt'SAFDL is.-caused by.asingle full strength C, EAdropiith CEACS, I and either a twelve finger or .Subgroup 'drop withCEACS oOS. (continued) PALO VERDE UNITS 1,2,3 B 3.1,5'-4 REVISION52

CEA Alignment B 3.1.5 BASES APPLICABLE In the case of the full strength CEA drop, a prompt decrease SAFETY, ANALYSIS i.n core average power and a dfstortion in; radial power areI (continued) ihitia'lly produced,-which When conservatively coupled result in. local power and h'at flux increases, and a decrease in DNBR.' A part.strength CEA drop would cause a similar .,reactivity response although with less of a magnitude due to the full strength CEAs having a more significant reactivity worth. WithCEACS OOS, a twelve finger and subgroup drop will result in greater 'radial power-distortion. To accommodate the greater distortion'without a reactor trip, increased -thermal margi.n is required to bepreserved. With CEACS IS,.as the-twelve finger drop is detected, core power and an appropriately augmented power distribution penalty 'factor are supplied to the CPCs. CPCs will trip if required to-prevent SAFDLs from being exceeded. For plant operation within the:DNBR and local power density (LPD) LCOs,,.DNBR and LPD trips can normally be avoided on a dropped 4-finger CEA sinceŽCEACs donot penalize DNBR or LPD for 6'.'fouffI finger'CEA*"drop. With CEAC. IS and.a subgroup drop, a distortion in power d:istrilbuti.on, ai'da decrease in core' power are produced. As .', ptheps~itionof the dropped'CEA subgroup is detected, an approprizate..power di stribution, penalty'factor is supplied to ..the CPCs, and a reactor' trip signal.qn low DNBR is generated.' -CEA'~ignment satisfies8 Criteria 2 and 3,bf 10 CFR ."50*301 )(2)( *) .i 0 LCO ThIe i mi~ts o1prt strength',-sh'downt and regulating CEA ý-'ý:alignments ensure that-:the;.assUmptions in the safety -,analysis-.wi.l'lremain valid' The requirements on OPERABILITY ensure that upon reactor trip, the CEAs will be available and will be inserted to provi'de enough negative reactivity to shut down the reactor. The OPERABILITY requirements also ensure that the CEA banks maintain the correct power distribution and CEA alignment. S..(continued) PALO-VERDE UNITS 1,2,3 B 3.1. 5-5 REVISION 52

CEA Alignment B 3.1.5 BASES (continued) LCO The requirement is to maintain the CEA alignment to within (continued 6.6 inches between any CEA and all other CEAs in its group. Failure to meet the: requirements of this LCO may produce unacceptable powerpeaking factors, DNBR, and LHRs, or unacceptable SDMs, all of whichlmay constitute initial conditions inconsistent with the safety analysis. APPLICABI LITY The.requirements on.,CEA OPERABILITY and alignment are applicable in-MODES 1 and 2 because these are the only MODES in which heutron (or.-fission) power is generated, and the OPERABILITY (e,.g., trippability) and alignment of CEAs have the potential to affect the,.s:afety of,.theýplant. In MODES 3, 4,*5, and 6,.the alignment.l1imits do not apply because the 'reactor is shut down'.and not-producing fission power. -In the:shutdown modes, the OPERABILITY of the shutdown and regul-ating,.:CEAs, h~as t.the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentrationof -the RCS. See .LCO0.3. 1" ",SHUTDOWN; MARGIN:-(SDM),ý., Reac,tor Tri p Breakers Closed,"" for SDM.i~n MODES 3,4, and-5,*and LCO 3.9.1, "Boron Concentration." for boron-concentration requirements during refueling. ~1 (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-6 REVISION 52

CEA Alignment B 3.1.5 BASES (continued) .ACTIONS A.-1 and A.2. A CEA may become misaligned, yet remain trippable. In this condition, the CEAcan still perform its required function of adding negative reactivity should'a reactor trip be necessary. If one or more CEAs (regulating, shutdown, or part strength) -are misaligned by 6.6 inches and ! 9.9 inches but trippable, or one CEA misaligned by > 9.9 inches but trippable, continued operation.in MODES,1 and 2 may continue, provided. within 1,hour, the power-is reducedin 'accordance with the limits in the COLR, and within 2 hour.s'CEA alignment is restored. Regulating and part strength CEA alignment can be restored by-either ali.gning.the misaligned CEA(s) to within 6.6 inches of itsgroup or aligning the misaligned CEA's groUp to within -6.6 inches of the misaligned CEA(s). Shutdbwn CEA.alignment can be irestored by aligning the .mlisaligned CEA(*)4..to-within.6.6 inches, of its group. X6n'on redistribdtion in-the core starts to occur as soon as a1CEA becomes'Imisal.ig'ned.:. Reducihg THERMAL POWER in accordance with"'the.limits in the COLR:ensures acceptable - power distributi~ons'are maintained.(Ref.;3). For small misalignments (< 9.9 inches) of the CEAs, there is:

a.

A small effect on the time dependent long term power distributions relative to those used in generating LCOs and limiting safety system settings (LSSS) setpoints;

b.

A negligible effecton the available SDM; and

c.

A small effect on the. ejected CEA worth used in the accident analysis. With a large CEA misalignment (Ž 9.9 inches), however, this misalignment would cause distortion of the core power distribution. This distortion may, in turn, have a significant effect on the time dependent, long term power distributions relative to those used in generating LCOs and LSSS setpoints. The effect on the available SDM and the ejected CEA worth used in the accident analysis remain small.. Therefore, this condition is limited to the single CEA misalignment, while still allowing 2 hours for recovery. (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-7 REVISION 52

CEA Alignment B 3.1.5 BASES ACTIONS A.1 and A.2 (continued) In both cases,, a 2 hour time period, is sufficient to:

a.

Identify~cause of a misaligned CEA:

b. Take appropriate corrective action to realign the CEAs; and

c.

Minimize the effects of xenon redistribution. The CEA must be returned to OPERABLE status within 2 hours. If a CEA misalignment results in the COLSS programs being declared INOPERABLE, refer to Section 3.2 Power Distribution Limits for applicable..actions. B.1 and B.2 At least.two of the..following.ihree CEA position indicator channels shall be.OPERABLE-fo.each CEA: y'a. CEA Reed Switch Position'Transmitter (RSPT 1) with,the c.aPabiIity,:lof determining the absolute CEA positions wi-thi'n 5.2 inches, CEA.-Reed 'Switch Position Transmitter (RSPT 2) wi th the2 capabi 1 ity..bf determiini ng the absolute CEA positions withi;n 5.2 inches, and c'. riTheCEA pulse.,countdinrg-posit-ion indicator .' ' '+.... channel

• u.

If only.one& CEA. position., indicator channel is OPERABLE for one CEA per CEA Group, continued operation in MODES 1 and 2 may continue, provided, within 6 hoursf at least two position indicator channels are returned to OPERABLE status; or within. 6_hoursand once per,.12 hours,.verify that the CEA . ' group with:the inoperable~positiion indicators are either ?uflly wi.thdrawn or.. fully inserted whii'aintaining the

• insertion limits ofLCO 3.1.6, LCO 31*;7and LCO 3.1.8.

CEAs are filly wi.thdrawn when therequti.rkments of LCO 3.1.6 +'~and 3.1,.'7 are met.',... -Addi.tionali]y;-.the Upper,Electrical Limit (UEL) CEA reed switches,;provide'an acceptable.,indication of CEA position for a fully, withdrawn conditioni. (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-8 REVISION 52

CEA Alignment B 3.1.5 BASES ACTIONS C.1 If a Required Action or associated Completion Time of Condition Aor Condition B is'not met, or if one or more regulating or shutdown CEAs are'untrippable (immovable as a result of excessive friction or mechanical interference or known to be untrippable), the unit is required to be brought to MODE 3. By being brought.to MODE 3, the unit is brought outside its MODE of applicability. When a Required Action cannot be completed within the required Completion Time; a controlled shutdown should be commenced. The allowed Completion Time of 6 hours is reasonable,-based on operating experience, for reaching MODE 3 from fullipower conditions in an orderly manner and without challenging plant systems. If a full strength CEA is untrippable, it is not available for reactivity insertion during a reactor trip. With an untripp~ab~le CEA,_meeting the ihs~ertibh limits of LCO 3.1.6. "Shutdowh'Control Element"Assembly (CEA)' Insertion Limits," and LCO 3.1.7,.".Reg.ulating Control Element Assembly (CEA) ,.. Insertion Limits,. does not ensure that adequate SDM exists. Therefore,. the pplahntmiust be'shut down in order to evaluate 'the SDM required boron concentration and power level for critical, operation, Continuedoperation is allowed with - - :untrippable part strength.CEAs if the alignment and insertion :lrimits.;are met:..

• Continued operation..isnot allowed with one or more full length CEAs untrippable., Thisis because these cases are indicative of a loss of SDM and power distribution, and a l.:

'oss of 'sa7fe~ty'functiOn,, respectively. .Cof*tinued -.operation is "not a'I'llowed 'ri ' the case of more than ({ne C'E.A misaligned from any'other CEAin-its group by '9>9, inches. group misaligned ,fromiany other. CEA&j in tthat. group by > 9.9. inches, or more than one CEA group that has a2-l]east ofie CEA misaligned from any other CEA in that group by > 9.9 inches. This is indic*tiveof..a'lossqo:f power'Ldistribu~ti.on and a loss of safety function-reSpectively.Mu-l.tipe CEA misalignments should result in'aitomatic pirotective action. Therefore, with two or more CEAs misaligned more than 9.9 inches, this (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-9 REVISION 52

CEA Alignment B 3.1.5 BASES ACTIONS D.1 (continued) could result in a situation outside the design basis and immediate action would be required to prevent any potential fuel damage. Immediately opening. the reactor trip breakers minimizes these effects.' SURVEILLANCE REQUIREMENTS SR 3.1.5.1 !Verification.that indiyidual CEA positions are within 6.6 inches'(.indicated reed'switch positions) of all other CEAsin the group at a,12.-1our Frequency allows the operator to detect a CEA that is beginning to. deviate from its expected position. The specified Frequency takes into account other CEA,.position information that is continuously available to the operator in the controlhroom, so that during actual CEA motion,'deviatidns can:,immediately be

detected, I

-_i SR 3.'.1 5.2' ,ýI;; OPERABILITY of at.ieast.two CEA position indicator channels is requir.ed to determine CEA positions, and thereby ensure compliance, with; the CEA" alignment'and insertion limits. The CEA full in and ful"ouit limits provide an additional independent means f-for determining the CEA positions when the .. CEAs :are. at. either their fully inserted or fully withdrawn .positions. SR 3.1.5.3 Verifying each-,ful.,T strength CEA iY

s. t pable would require that eachCEA be.tripped.

In,

• MpES:1.,arhd 2 tripping each ful I ",!streng,th,-.CEA,woul'd resu'l-tL 'I h.. n

,adi -l or axial power tilts, or,,oscillations.,Therefore indi'vidual full strength CEAs ar'eeercised-every.;92, days to iprovii de increased rconfidence.thCatall full s,trength-'CEAs:r axtinue to be "trippabld, eVen "if they are rnot regularly; tripped. A imovement."of' 5 iinches<is adequate"to, demdnstrate motion without exceedi rg""thealign"mentilimrif",Whe.n only one full strengthiCEA is being moved,". The'92.'.djy.Frequency takes into c6drsideration._other informati.on' 2available to the operator in the cohtrol room and other surveillances being performed more frequently.,which add to the determination of OPERABILITY of the CEAs (Ref. 3). Between required (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-10 REVISION 52

CEA Alignment B 3.1.5 BASES SURVEILLANCE REQUIREMENTS SR

3. 1.

5.3 (continued) performances of SR 3.1.5.3, ifa CEA(s):yi~s discovered to be -immovable but remains~trippable and aligned. the CEA is .considered to be OPERABLE. At anytime, if a CEA(s) is immovable, a determination of the trippability (OPERABILITY) of that CEA(s) must be'made, and appropriate action taken. SR 3-:1.5.4 Performance of a CHANNEL FUNCTIONALTEST of each`reed switch' position transmi.tter.channel ensures the channel is OPERABLE and capable of.indicating CEA position. Since this test ,must be performed When the reactor 'is shut down, an 18 month .Frequency to,becoi.ncident with refueling outage. was selected. Operati'ng experience has shown that these u,components ubally pass this Surveillance when performed at !a Frequencyi.of once evey.!18 months.. Furthermore, the Frequency takesinto account other factors; which determine S.the OPERABILITYof the.CEA Reed Switch Indication System. These factors ihclude:

a.

Other, more frequently performeo..surveillances that help to verify OPERABILITY;..

b.

On i.6ne6 diagnostics performed automatically b the CPCs', CEACs, and the Pl~ant Computer which include CEA positi-on comparisons and sensor validation; and v! : ,y.

c.*

The CHANNEL CALIBRATIONs for the. CPCs (SR 3.3.1.9) and " CEACs (SR 3.3;.3,.4) inpuLt channels.'that are performed at 18 month intervals and is,..an ove~r~lapping test. SR 3.1.5.5 Verification of full strength CEA drop times determines that themaximum-CEA 'drop time permitted ds consistent with the as§uimed -drop timeluse.Ji'n'the'*gafety ana:lysis (Ref. 3). Measuring idrop times prior,.to rleac~tor criticality, after

re actor'vessel head remov~al,:,ensures the reactor internals

• .,:and CEDM will not iinterfererwi.th CEA motion or drop time,

!.!and that no degradation in these-systems has occurred that would adversely affect CEA motion or.drop time. Individual GEAs-,Whosedrop:.times'ane greater than safety analysis .a,ssumptions:'are.not OPERABLE.. This SRis performed'prior to '7.!.;ciiticalitydue~to t0e* yant conditions needed to perform Sthe SR and the,pdtenti~a T!fr an unplanned plant transient if ....,the Surv'eillance we~e'.performed wi~th the reactor at power. (continued) PALO VERDE UNITS 1,2,3 B 3.1..5-11 REVISION 52

CEA Alignment B 3.1.5 BASES SURVEILLANCE REQUIREMENTS SR 3.1.5.5 (continued) The 4 second CEA drop time is the maximum time it takes.for* a fully withdrawn:individual full strength CEA to-reach its, 90% insertion position when electrical power is interrupted to the CEA drive mechanissm with RCS Tcold greater than or-equal.to 550°F and all'reactor coolant pumps operating. The CEA drop. time of full'strength CEAs shall also be demonstrated through.measurement prior to reactor criticality for speci'fically affected individual CEAs following any maintenanceion or modification to the CEA drive system. which could'affect the-drop time of those specific CEAs, REFERENCES 1., 10 CFR 50,.Appendix.A. GDC 10 and GDC 26.

2.

10 CFR 50: 46. ..3. 'UFSAR, Section: 15.4.

4.

UFSAR, Section, 7. 7.1.3. 2. 3

5 UFSAR,. Section 7, 5.

1.4 PALO, VED, NT 3151 EIIN5 PALO VERDE UNITS 1,2,3 B 3.1.5-12 REVISI'ON52

Part Strength CEA Insertion Limits B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1'.8 Part Strength Control Element Assembly (CEA) Insertion Limits BASES I BACKGROUND The insertion limits of the part strength CEAs are initial assumptions. in the safety analyses for CEA misoperation events.' The insertion limits'directly affect core power distributions. The applicable criteria for these power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design"' (Ref. 1)...and 10CFR 50.46, .. "Acceptance Criteriajfor Emergency Core Cooling Systems for

Light Water NuclearPlants"_'(Ref.-2).

Limits on part strength CEA insertion have'been established, and all CEA positions are monitored and controlled during power operation to ensure that the power distribution defined by the design power peaking limits is preserved. 'The part-'strength CEAs are 'used for axial-power shape control of the reactor., The positions of the part strength CEAs are manually controlled. They are capable Of changing reactivity very quickly..(compared.to borating or diluting). The power.density-at any point in the core must be limited to maintain specified acceptable fuel design limits, including limits thatpreserve'thecriteria specified in 10 CFR 50..46.(Ref-,...2)::. Together,..;LCO. 3.1.:7, "Regulating Control Element Assembly (CEA) Insertion Limits"': LCO 3.1.8; LCO 3.2.4, "Departure From Nucleate Boiling Ratio (DNBR)"; and LCO 3.2.5, "AXIAL SHAPE INDEX (ASI)," provide limits on control component operation and on monitored process variables to ensure the core operates within the linear heat rate (LHR) (LCO 3.2.1, "Linear Heat Rate (LHR)"); planar peaking factor (F y) (LCO03.2.2, "Planar Radial Peaking Factors (Fy)"): and LCO 3.2.4 limits in the COLR. Operation within the limits given in the COLR prevents power peaks that would exceed the loss of coolant accident (LOCA) limits derived by the Emergency Core Cooling Systems analysis. Operation within the Fy and departure from nucleate boiling (DNB) limits given in the COLR prevents DNB during a loss of forced reactor coolant flow accident. (continued) .,PALO VERDE UNITS 1,2,3 REVISION 52

Part Strength CEA Insertion Limits B 3.1.8 BASES BACKGROUND The establishment of limiting safety system settings and (continued) LCOs requires that the expected long and short term behavior of the radial peaking factors be determined. The long term behavior relates to the variation of the steady state radial peaking factors with core burnup:' it is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion islassumed, and the expected power level variation throughout'the cycle. The short term behavior.relates to transient perturbations to.the steady state radial: peaks due to radial xenon 'redistribution. The magnitudes of such perturbati'ons depend upon the expected use.of.the CEAs.during anticipated power reductions and load maneuvering. Analyses' are performed,-'based on the expected mode ofoperation of the Nuclear Steam Supply System (base loaded, maneuvering, etc.). :From these analyses, CEA insertions are determined, and a consistent Set of radial peaking factors are defined. The long term (steady state) and short term insertionlimits-are determined, based upon '" the assUmed mode of oper'ation used,.t'.the"analyses: they provide a means ofp'Dese'rving.the a6sý6nptions on CEA insertions' used>. The longý'and sh6ort term'.insertion limits of LCO 3.1.8 are specified for the plant, which has been designed primarily-tor base loadedoperation, but has--the..... ability,.to accommodate a limited.amount of load maneuvering. APPLICABLE SAFETY ANALYSES The fuel cladding must not sustain damage as a result of rnormal operation -(Condition I) and anticipated operational - occurrences.,(Conditi.on II),. The regulating CEA insertion, hart ste.n'gth CEA, inseftion, 'ASI, 'and Tq`LCOs precl'ude 'core.-- power di-stributiohs fromz'occurring',that would violate the followi'ng fuel deýsign' criteri 'a.

a. 'Dur ihg ai:a rge break LOCA>" the. pea.k cladding temperature must not"excedd 22006F.(Ref. 2) b.. During CEA misoperation events, there must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition;

c.

During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 3); and (cohtlh0ed) PALO VERDE UNITS 1,2,3 B 3. 1.8-2 REVISION 52

Part Strength CEA Insertion Limits B 3.1.8 BASES APPLICABLE SAFETY. ANALYSES (continued)

d.

The CEAs must be capable of shutting down. the.reactor with a minimum required SDM, with the highest worth CEA stuck fully withdrawn. GDC 26 (Ref. 1). Regulating CEApositi~on, part strength CEA position, ASI, and..Tq are process variables that together characterize and control'the three dimensional power distribution of the reactor core.,. Fuel cladding damage does not occur' when 'the core is operated Outsi.de these LCOs during-normal operation. However, fuel cladding damage.could ýresult, should an accident occur-wi.th simultaneous-'violation of one or more of these LCOs.; Changes in.the power distribution can cause increased power peaking and corresponding increased local ILHRs. ,,The part strengthCEA. insertion limits, satisfy Criterion 2 of i10.CFR 50.36 (c)'(2)(ii).. The part strength CEAs are required'due to the potential peaking factor violations that could occur if part'strength CEAs: exceed.insertion limits. LCO* The limits on part strength CEA insertion, as defined in the COLR. must be maintained because they serve the function of preserving powerdistribution. APPLIICA 'IITY Th6e part s.itrengthj j nser.tion limits shall-be maintained with t.he.,r heacto.::.i n '.MODES:,1 and -2 Thesel imits must be maitairid,.since they presere. the assum ed power di stri buti on. Appl :icabflity I lower MODES is not required. , si'nce the power distribution assumptions would not be .. xc'6eeed i,*,these(MODESn ue (continued) PALO VERDE UNITS 1,2.3 . B 3-1. 8 -3 REVISION 52

Part Strength CEA Insertion Limits B 3.1.8 BASES (continued) ACTIONS A.1, A.2 and B.1 If the;part strength CEA groups are inserted beyond the following limits, flux patterns begin to develop that are outside'the range assumed, for lon' term fuel burnup;,

1)

Transient insertion limits: ,2) Between the long-term (steady-state) insertion limit and the transient insertion limit for: a) 7 or more effective full power days (EFPD) out of any 30 EFPD-period: b) 14 EFPDor more out of any 365-EFPD period6..,. :. If allowed to continue beyond this limit,"the peaking factors assumed as initial conditions in the-accident analysis may-be invalidated (Ref.'4): Restoring the CEAs to within limits or.reducing THERMAL POWER to that fraction of RTP that is allowed by CEA group'position, using the limits specified in the COLR, ensures that acceptable peaking -factors are-maintained.- Since these effects are cumulative, actions are provided to limit the total time the part strength CEAs can be out of limits in any 30 EFPD or 365 EFPD period. Since the cumulative out of limit times are in days, an additional Completion Time of 2 hours is reasonable for restoring the part strength CEAs to within the allowed limits. C.1 When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should commence. A Completion Time of 6 hours is reasonable, based on operating experience, for reaching Mode 3 from full power conditions in an orderly manner and withoutchallenging plant systems. (continued) PALO VERDE UNITS 1,2,3 B 3.1.8-4 , REVISION-52

Part Strength CEA Insertion Limits, B 3.1.8 BASES (continued) SURVEILLANCE REQUIREMENTS SR 3.1.8.1 Verification of eachpart strength.CEA group position every 12 hours. is sufficient to detect CEA positions that may

approach the limits' and provide the operator with time to undertake the Required Action(s), should insertion limits be found to be exceeded.

The 12 hour frequency also takes into account the indication provided by the power dependent insertion limit alarm circuit and other information about CEA group positions available to the operator in the control room. REFERENCES

1.

.0 CFR,50. Appendix A, GDC 10 and GDC 26. -2. .10 CFR*5O..46.,

3.

Regulatory Guide 1,77,"Rev. 0. MaY 1974. 4.. UF:.SAR, Section 15.4.. PALO VERDE UNITS 1,2133 B 3.1.8-5 REVISION 52

This page intentionally blank:.. F1 ? I'. I, 4'

F I B 3.2,ý B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Planar Radial Peaking Factors (FY) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Coolant Accident (LOCA), loss of flow accident, ejected Control Element Assembly (CEA) accident, or other postulated accident requiring termination by a Reactor Protective System (RPS) trip function. This LCO limits damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset: bf a'transient. Methods of controlling the power distribution include:

a.

Using full strength or part strength CEAs to alter the axial power distribution;

b.

Decreasing CEA insertion by boration, thereby improving the radial power distribution; and

c.

Correcting off optimum conditions (e.g., a CEA drop or misoperation of the unit) that cause margin degradations. The core power distribution is controlled so that, in conjunction with other core operating parameters (CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. Limiting safety system settings and this LCO are based on the accident analyses (Refs. 1 and 2), so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs), and the limits of acceptable consequences are not exceeded for other postulated accidents. Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor in controlling axial power distribution. Power distribution is a product of multiple parameters, various combinations of (continued) PALO VERDE UNITS 1,2B3 B 3.2.2-1 REVISION 52

F B 3.22. BASES BACKGROUND whichmay produce acceptable power dist'ributions. Operation (continued) withji.n the design limits of power distribution is accomplishediby generating operatinglimits on Linear Heat Rate.(LHR) and Departure from NucleateBoiling (DNB). Proximity to the DNB condition is expressed by the Departure from Nucleate Boiling Ratio (DNBR),*defined as the ratio of .the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux. Theminimum DNBR value during both'normal operation and' AOOs is! the DNBR.Safety Limit as.calculated by the,CE-1 Correlation (Ref. 3) and corrected for such factors. as rod bow and grid spacers, and it is accepted as an appropriate margin to DNB for all operating conditions There are two systems that monitor'core power distribution online: the. Core Operating"Limit Supervisory System (COLSS) andthe Core'.Prdtection'Calcul..ators"(CCPCs). The COLSS and CPCs that monfitor the4core power distribution are capable of verifying thýt the LHR and"-.the-DNBR-do not exceed their limits. The COLSS.jperforms this functibn by continuously monitoring the.cdre dpoer distribution'and calculating core poweroperating limits corresponding tothe allowable peak LHR *and DNBRkvalue... Th6 CPCsy..performlthis function by cqntinuously ca.luiatingactual values',of DNBR and Local Power Den'sity (LPD) for comparison'with the respective trip setpoints. DRpenalty factors are included in both the COLSS -and CPC .,DNBR calcuilations to,a commddate the effects of rod bow. The amount 6f r'od bow 'in'e'ach assembiybI dependent upon the average..burnup experienced by that assembly. Fuel assemblies that incdurhigher than average burnup experience greater-.rod bow. Conversely. fuel assemblies that receive ..lower thanaverage.unup exper'ience less rod bow.n d'design, cal~cul'atidn.for a relload core,*;each batch'f 'fuelis ass igned',a penalty appliedto the'maximum integrated planar radialpower.peak of the' batch;.'T's-TSehalty is correlated with the amount of rod bow determine'd from the maximum average.assembly,burnup of the batch.. A single net penalty for the"COLSS'andCPCs is then determined from the penalties associated with each batch that comprises a core reload, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches. The COLSS indicates continuously to the operator how far the core is to the operating limits and provides an audible (continued) PALO VERDE UNITS 1,2,3 B 3.2.2-2 'REVISION 10

F B.. BASES BACKGROUND (continued) alarm if an operating limit is exceeded. Such a condition signifies a reduction in the capability of the.plant to withstand an anticipated transient, but does not-necessarily imply an' immediate violation of fuel design limits. If the margin to fuel design limits continues`to decrease, the RPS , ensures that the.specified acceptable fuel design limits are not exceeded for.AOOs by initiating a'reactor trip. The COLSS continually generates an assessment of the calculated margin for LHR and ODNBR specified limits. The data required for these assessments include measured incore neutron flux, CEA positions, and Reactor Coolant System (RCS) inlet temperaturie, pressure, 'and flow. In addition to monitoring performed by the COLSS, the RPS (via theCPCs).continually infers the core power distribution' and' thermý margins by processing reactor Icoolart data,.sgnal'from excore neutron' flux detectors, -,ad input from redundant reed switch assemblies that 'indicates CEAp6osition'. In this case, the CPCs assume a ,ninimum:.core powqer..of20% RTP,. This threshold is set at 20% RTP bedause'the. power range excore'neutron flux 'detecting system. Is inaccurate below thi's power level. If 'power-distribution or other parameters are perturbed as a result of an AO0; the. high'. LPD or low DNBR trips in the RPS

initiate a reactor. trip prior t6oexceed.ing fuel design limits.

The'limits. on.ASI, FX,, and.3q represept.limits within which theL'HRa andDNBR algorithms are' t.Va:id."hese limits are btaineddirectlyfr6mthe ini.i.al core'or reload analysis. APPLICABLE.;, SAFETTY ANALYSES The fuel ",iladdi ng must not susýtain damage as a result of .or. mal operation'or, AOOls "(Ref. '4).;. The power distribution .and CEA insertion and ali "rimen'tLCOs prevent core power distributions from reaching'l evels' thae followi ng fuel design criteria; ..a,.. Dung.a LOCA,-peak. c dding temperature must not G.' " i. exc.ee .d ']'22 0 _F (k f,'".,. :5 u.. (continued) PALO VERDE UNITS 1,2,3 B, 3. 2. 2-3 REVISION 0

F.' B.. BASES I APPLICABLE SAFETY ANALYSES (continued)

b. During CEA misoperation events or a loss of 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 DNB condition.:(Ref. 4):

c.C. During an ejected CEA accident, the.fission energy ainputdto the fuel must not exceed 280 cal/gm (Ref. 6); and

d.

The control rods (excluding part strength rods) must be capable of shutting down the reactor with a minimum required SDM with the highestworth control rod stuck fully withdrawn (Ref. 7). The power density at any pointin the core must be limited to maintain the fuel-design criteria (Refs.- 4 and 5).- This result is accomplished by maihita'iriing the power distribution and reactor coolant,-conditions so that; the peak LHR and DNB parameters are withfn 6perating limits s'uipported by the accident analyses..(Ref. 1) withdue. regard for the correlations 'between measured quantities, the power distribution, an~d theuuhce'rtainties in the determination of power.,distri~buti on. Fuel 'cladding failur6dbiring'a LOCA is limited by - restrictinq.the maximum Linear Heat Generation Rate (LHGR) so that thepeak 'ladding temperature dobs not exceed 2200'F (Ref. 5). Peak cladding temperatures exceeding 2200°F cause severe cladding failure by 6xidation due to a Zircal-oy water -. reaction, The LCOs" gbvern.ing. LAR,:ASI,..CEAs,, andRCS ensure that these criteria are'.et a~s'l'6ng',s the'core.. iJs operated within the

ASI.,and.F " IlImits.specified n.-the COrR'-

and within the Tq l J imits. 1,,#he latter,are prdcess variables that characterize the_ three, d imerisional power" d,stribution'of the reactor core'. Operation within. the limi~ts for these variables ,ensures tha'ttheir actual, values are-within the ranges used in the atcident analyses (Ref. 1). "' i 7. Fuel cladingdamage does not occur because of conditions outsidethe,.limi tsof these LCOs,'forA'S.I'. Fy, and Tq during normal operatidn. -However ':fue cladding damage results if an accident occurs from 'initial 'conditions outside the limits of these LCOs. This potential for fuel cladding damage exists because changes in the power distribution can (continued) PALO VERDE UNITS 1,2,3 B 3.2.2-4 REVISiON'52 lk

B 3.2. B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AZIMUTHAL POWER TILT (Tq) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO either limits or prevents potential fuel-cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Coolant Accident (LOCA), loss of flow accident, ejectedCControl Element Assembly (CEA) accident, or other postulated accident requiring termination by a Reactor Protective System (RPS) trip function. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a transient. "Method's.of controlling th power distribution include: a: .Us'! ful'stre'gth or'part strength CEAs to alter the ..aial power. distributin;-

b.

Decreasing CEA ins'ertion' by bbra`tibh. thereby improving the radial power distribution: and c... Correcting' off-optimum conditions,'(e.g., a CEA drop eOr misopoeratn .f.the unit) that cause margin, The core power distribution is coh'tro61led so that, in ,conjunction with. othier core operating parameters (e.g., CEA ins'er'tion nd an'ignment lim`-ts)'. the oower distribution does 'not.resl t in violatioh of thi L'CO'- 'The limiting safety system.settings and this LCO are based-bn the accident analy'ss.,(Refs ',1 and *2), 'so 'that specified acceptable fuel design l imi ts are :not 'exceded as, a result of Anticipated Operational Occurrences (ADOs)'and the'.limits of acceptable consequences are not exceeded Ofr other postulated accidents. -imiting'power dIisribution -skewing over time also minimizes xenon distribution skewing, which.is -a-significant factor in 'Qontroil I i ng axia1 power di str.-i buti on". (continued) PALOQVERDE UNITS 1,2,3 B.3.2.3-1 REVISION 52

B 3.2. BASES BACKGROUND Power distribution is a product of multiple parameters'," (continued) various combinations-of whi'ch may produce acceptable power. distribbutions. Operation within the design limits of~power distribution is accomplished'tby generating operating limits on the Linear'Heat Rateý (LHR) and the Departure from Nucleate.Boiling (DNB):. Proximity.to the DNB condition :is expressed by the Departure from Nucleate Boiling Ratio (DNBR), defi.ned as the ratio of the cladding surface heat flux required to cause DNB to the actual cladding*sdrface heat flux. The mithimum DNBR value duringboth!,normal operation and'AOOs is the DNBR Safety Limfitas calculated bythe CE-1 Corrielation (Ref. 3) and corrected for~suchfactorsas rod bow andgrid spacers, and it is accepted~as an approoriate"margin'. to DNB for all operating condi ti ons. There:are two systems that inoni, tor core power distribution online: -the Core Operating Limit-Supervisory System (COLSS) a'nd the Core Protectioh Calculators (CPCs). The COLSS and -CCs'.that~monitor the-. core power distribotion are capable of evrifying that the LHR, ahd-the DNBR'donot exceed their Tlimits. The COLSSperfor;'ms thi's.function tby continuously moni tori rig the, cdreower,,distr.ibtuion ahd calculating core power operating limits Corresponding'to-the allowable peak LHR and DNBR. The CPCs performthi's function by ..ontinudusly calculating aftual values of DNBR and Local Power' Density (RLPD.)-forcomparisop withW"the respective trip .setpoints. A'.DNBR enaltS' factor*--S :.ihcluded, in:the COLSS and CPC DNBR . calcul.ati6n"nto~accomm6date-the effects~oflrod bow. The amount 'If rod bow in dac.h'asseth bly-i*s dependent upon the average burnup experienced by the sasembly. Fuel assemblies that incur higher than average burnup experience greater magnitude-of rod-bow..- Conversely,.fuel. assembliesthat. -. receive lower than average burnup experience less rod bow. In desi gn`aldulations for arel'oadýcore:,f each batchl'of fuel ,is assigned a penalty appli'ed tot;hemaximumin~tegrated. planar,-radial power peak of the batch., This penalty is ".correatd with'the-.amount of rod -bow.,that is determined from the maximum average assembly burnup-of the batch. A single net. penalty for the COLSS and CPCs is then determined from the-penaties 'asocated with.each batch that comprises a core reload,ýýaccobnting.fo'r the-'offsetting margins caused by the lower radial power peaks in the higher burnup batches. (continued) PALO VERDE UNITS 1,2,3 B 3.2.3-2 RE-VISION.10

T3 B 3.2.3I BASES BACKGROUND.- (continued) The'COLSS. indicates continuously to the operator how far the core is from the operating limits and provides an audible

alarm if an operating limit is exceeded.

Such a condition signifies a reduction in the capability of the plant to Withstand an.,anticipated transient, but does not necessarily imply an immediate violation of fuel: design limits. If the margin to fuel.design limits continues to decrease, the RPS ýens.ures that the specified:acceptable fuel design limits are 'not exceeded for. AOOs by initiating a reactor trip. The COLSS continually generates.an-.assessment of the ,,calculated marginjfor LHR and DN'BR.,specified limits. The data required for these assessmeints include measured incore neutron flux data;, CEA positions, and Reactor Coolant System (RCS).inlet temperature, press~uree, and flow. In addition to the monitoring performed by the COLSS, the RPS,(via thef..PCs)..continually infers the core power .,,distribution and..thermal margins.byprocessing reactor. cool.ant~data.signals from excore neutron,flux detectors, !,;and input from redundant reed switch assemblies that

a.

indicates CEA.position..In this case, the CPCs assume a .'mi~nimumcore,*power of 2?0%.RTP. This.,threshold is set at 20% RTP because'the power~range excore neutron flux Adetection system is i1naccurate below-this power level. If .,,power distribution or other parameters *are perturbed as a

result of. an AOO., th!e higoh local power, density or low DNBR

.trips in the RPS -initiate a reacto r tripprior to exceeding fuel design limits.

.Thel imi ts.:on the ASI., F and TI represent limits within which -the,.LHR, and:,DNBI3 a 'o rthm s,are val i d. These limits rae. obtaiined.directly*romth*e.' ial re or reload nalsis. ""'-CAB.. E, Th "f e cl, ad in ',mus )

.t *,,

APPLICABLE'> , The fuel, csaddln 'mus t.notsustain` damage as a result of SAFETY ANALYSES

Operationand AOOs (Ref..4).

The'power distribution and CEA inserti'o*rr and alIignment LCOs pr6clude core power .-distributions, that vi'olate "the following fuel design cri~teria:. ý .Dring a LOCA..peak claddilbg.'tem'erature must not "-.exceed 2200.F- (Ref..5); (continued) PALO.-VERDE UNITS 1,2,3 B. 3.2.3-3 REVISION 0

B 3.2.3 BASES I APPLICABLE SAFETY ANALYSES (continued)

b.

During CEA misoperation events or a loss of 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.DNB condition (Ref. 4);

c.

During a CEA ejection accident, thefission energy-input to the fuel. must not exceed 280.cal/gm (Ref. 6): and

d.

The controlI rods (excludi!ng part strength rods) must be capable of-shutting.down the reactor with a minimum required SDM with the highest worth-control rod stuck fully withdrawn (Ref..7,). The power density at any point in the..cQre must be limited Jto-maintainthe fuel design criteria-(Ref. 1). This result is accomplished by maintaining the.,powerdistribution and reactor coolant conditions so that the peak LHR and DNB parameters are within'operating limits'supported by the accident analysis (Ref. 2) with due regard for the corre-lations-between measured quantities,, the power distribution, anduncertainties in the determination of. power distribution.

Fuel 'cladding'failure during a.LOCA, is 1i mi ted by restricting the maximum Linear Heat Generation Rate (LHGR)

-so that thet.peak Cladding temperature does not exceed 2200°F (Ref. 1.*). Peak cltadding~temper atures exceeding 2200°F cause severe cladding failure.byoxidation due to a Zircaloy water reaction. t .I. The LCOs.governi:ng LHR, -ASI, CEAs;.,.ýjd -RCS ensure that these criteri'a are. met aslong as the cqre.is.operated within the ASI andFI* :,imits specified in the.COLR,-,,and within the Tq l i'mi ts he latter.',"are process -variables that characterize the three6dimensional power-distributioniof the reactor core. Operation within the limits of these variables . ensures that;their.:actualivaiues are within the range used in!the accident analyses (Ref.:.1) (continued) PALO VERDE UNITS 1,2,3 B 3.2.3-4 REVISION 52

DNBR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Departure from Nucleate Boiling Ratio (DNBR) BASES BACKGROU ND The purpose of this LCO is to limit.the core power distribution to the initial value assumed in the accident analyses. Specifically, operation *within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in:..the event of a Loss Of Coolant Accident (LOCA). loss.of flow accident, ejected. Control Element Assembly* '(CEA). accident, orother postulated, accident requiring termination by\\a Reactor Protective System (RPS) trip function. This LCO limits the amount of damage to the fuel claddi'ng during an accident by ensuring that the plant is operating :Within acceptable.conditions at the onset of a transient.' Methods' of control ling. the 'power-di stri bution include: a., -Using :ful.l strength.or part st*rength CEAs to alter the axial power distr.bution'

b.

Decreasing CEA insertion by boration, thereby " ;improving the radial power -distribution: and ... *.~

c.

Correcting offloptimum conditions (e.g., a CEA drop or mi-soperation.of"..the uni.t), that cause margin degradations.'.., 2: The core power distribution is controlled so that, in 'conj~unction..with'ý oth~er core'operating parameters (e.g., CEA '1 nsertibon and a 1 i gnmentl imi'ts), the power di stri buti on does 'not result i~n violatiton of: this.,.LCO,,. The limiting safety Syst'em'settings-and this,..LCO are based on the accident ,""analYsis (Refs...N.ands.2), -so ý;that.specified acceptable fuel .design -limits are not"exceeded as a result of Anticipated 'Operati onal;-Occurrences (AOOs)' :and?-.the-limits of acceptable consequences are not.exceeded"for other postulated accidents. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which isa significant factor in controlling axial power distribution. (continued) PALO VERDE UNITS 1,2,3 B 3.2.4-1 -REVISION 52

DNBR B 3.2.4 BASES BACKGROUND Power distribution is.a product of multiple parameters, (continued) Various combinations' of which may produce. acce0table power distributions. Operation. within-the design limits of power distribution is accomplished by generating operating limits on the Linear Heat Rate (LHR) and the Departure from nucleate boiling (DNB)" I Proximity,'lo the.DNBconditionis expressed by the DNBR, defined as the ratio of the cladding surface heat flux required-to cause DNB to the actual cladding surface heat fluý..*The minimum' DNBRvalue during both~normal operation and*AOOs js theDNBR Safety-;Limitas'calculated by the CE-1 Correlation (Ref. 3) and corrected for,ýsuch factors as rod bows and grid spacers and it is accepted as an appropriate margin'to DNB for all operating*conditions. There are two systems that monitor core~power-distribution online: theCore Op~rating. Limits Supervisory System (COLSS). and the Core Pr6tection-Ca'lculators (CPCs). The COLSS ahd CPCs that monitor' the core 'power distribution are capable'.bf verifying that. th'e,.LHR.and DNBR do not exceed their limits'. The COLSS perfOrmbs thi~s function'by continuously mohitoring the, core power distribution and Cal'ulating corepower~operating limits corresponding to the alldwab.le peak LHR'and 'DNBR:. The CPCs.perform this function by continuously calcula ting 'an ýactual 'value of DNBR and LPD for comparison with the respective trip`setpoints.

.A DNBR penalty factor. is.included in both the COLSS and CPC S DNBRI calculation to accommodate-'the effects of rod bow.

The amount of.rod.bow in,'achassembly isý dependent upon the average burnup ekperiehced by'tnat assembly. Fuel assemblies that incur higher than average burnup experience a greater magnitude-of rod bow. Conversely,-fueI.assembl.ies that,receive lower than average burnup experience less rod "",bow. r neSgnicaclationsfor a 'reload core, ech b*atch., of fuel 'is., assigned ý "ena tythat isf'applied tthe maximum integrated plana.rradiall power-peak.: f the batch. This penalty is correlated with the arn6unt obfrod bow that is determined Tr6m the'maximum average. assembly burnup of the 'batch. A,,single net penalty for the COLSS and CPCs is then determinhed'from the.penalties a"ssbciated:.with each batch that comprises-a-core reload. -accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches. (continued) PALO VERDE UNITS 1,2,3 B 3.2.4-2 REVISION -10

DNBR B 3.2.4 BASES BACKGROUND. (continued) The COLSS indicates continuously.to the operator how far the core is from the operatinglimits and provides an audible alarm when.an operating limit.is. exceeded. Such a condition signifi-es.a reduct ion in the capability of the plant to -withstand:an antici.pated transient,: but does not necessarily implyan immediate viOlation of fuel design limits. If the margin to fuel design limits continues to decrease, the RPS ensures,that.the_.specified acceptable fuel design limits are not exceeded during AO0s by initiating a reactor trip. -The COLSS continually generates an assessment of the -calculated margin for LHR~and.,DNBR specified limits. The data required for these assessments'include measured incore neutron flux, CEA positions, and Reactor..Coolant System (RCS) :inlettemper, ature, pressure, and flow. In addition to, the monitoring performed by the COLSS, the RPSj(v,*ia*he.'CPCsý);continually infers the core power " distnibution.and therml margins by processing reactor c~oolant, data,,siignals from excore, neutron flux detectors, and input' fromredunidant reed switdh'assemblies that indic ates, CEA-positioon'. 'In this case,,the CPCs assume a minimum. core..power of 20%.RTPbecause.the power range excore .neutro flux:detecting system'-iS inaccurate below this power ,level. If power distribution or other parameters are perturbed 'as a~result of an AGO. the high local power densityor lowDNBR.trips-in..th'eRPS, in.it'iate a reactor trip prior to exceeding fuel design limi'ts. ". The.limi ts on AS I, xFy' and.. T, rrepreseht limits within which .the LIHR. and.,DNBR, algorithms :re'va 1id.:. These limits are .obtained di.rectly-from, the

irii,

..core or reload analysis. APPLICABLE,: The f~fue'l c~ladding must not sustai n. damage as a result of

SAFETY ANALYSES.... normal operation,.or AOOs (Ref..4)..

The power distribution A. and.EA i nserft~i on, 'and al'ign'meht LCOs prevent core power .distributi ons. fromrn reachi ng. Tevel s. that'vi ol ate the So.i*, lng f 6 l,.design' cri te.iaa:. a Duri.ng a. LOCA,.peak cladding, temperature must not exceed 2.200 0 F (Ref. 5); (continued) 'PALO.VERDE UNITS 1.2,3 ,B 3.2.4-3 REVISION 0

DNBR B 3.2.4 BASES APPLICABLE

b.

During CEA misoperation events or a loss of flow accident, there must be at least SAFETY ANALYSES 95% probability at the 95% confidence. level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 3);

c.

During an ejected.CEA accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6): and

d.

The control rods (excluding part strength rods) must.be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawh (Ref. 7). The power density at any point in the c6oe must be limited to maintain the fuel design cri.teria(Ref. 4). This is accomplished by' maintaining the power distribution and reactor,.coolant conditions, so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses.(Ref.. 1) with due-regard for the correlations between'.measured quant'ities: the power distribution, and uncertainties in the determination of power distribution., Fuel cladding failure during a LOCA is limited by , restricting~the maximum Linear Heat-Generation Rate (LHGR) so,,tbhat th.e Peak. c ladd-ing-temp'erature does not exceed 2200°F (Ref. 4).' Peak cladding tempera'tures exceeding 2200°F may cause severe cladding failure by oxidation due to a Zircaloy .:water reaction.,,, The LCOs Egoverning LHR,i*ASI,, CEAs'. ndR ensure that these cr*iteria-are met aslong.as the cpoe..is.6perated within the A*I~~J aniF *lits-speci..fi1ed in 'h',CLR land within the T, limilts.,.he latter are-process variables, that characterize .the threeidimenesional power distibtition-of the reactor core. Operation,within the limfts.for these variables ensures that'. their actual val uesa, r-e., w'i t'i n the range used in the accident analyses (Ref. 1). i (cont~inued) PALO VERDE UNITS 1,2,3 8 3.2.4-4 REVISIOW52

ASI B 3.2.5 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.5 AXIAL SHAPE INDEX (ASI) BASES BACKGROUND The purpose of this LCO is to limit' the core power distribution to the initial values assumed in the accident analysis. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and. release fission products to the reactor coolant in the event of a Loss Of Coolant Accident (LOCA), loss of flow accident, -'ejected Control Element:Assembly (CEA) accident, or other postulated"accident requiring termination by a Reactor Protective System (RPS) trip function, This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditidns-'at theIonset of a transient. Methods Of "iontorolling the axifal power distribution include:

a.

Using full 'strength or part strengthWCEAs to alter the ' axial 'pow&er distribution;

b.

Decreasing CEA insertion -by boration, thereby improving the axial power distribution: and c Correcting offi'timum conditibns (ýe.g., a CEAdrop or ".mi soperation. ofthe. uhit.) that.:"cause margin degraa'datiiois..i The core power distribution is' dnhtrolled,*so that, in conjunction, with.other core operating parameters (CEA insertion and alignment YlimitS), the power distribution does .:.,t*ot:re'sUlt 'iriviolati on,:of'thifs :LCO.. The limiting safety 2

• sysitem setti'ngs are based on'- the accidentý analyses (Refs. I

' "aftd:2), "so that spdCifid-acceptable~fUel design limits are ..ot e'x'eed*d as a resul't-of Ahti c.ipated,.Operational cOCurrences"*(AOOs) aiid the limits:"of acceptable consequences are not exceeded *foh otherposi)Ulated accidents. Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor.in controlling axial power distribution. (continued) PALO VERDE UNITS 1,2,3 B 3.2.5-1 REV I S I ON 5.2

ASI B 3,2.5 BASES BACKGROUND Power distribution is a product,.*of multiple parameters,_ (continued) various combinations ofwhich may produce'acceptablepower distributions. Operation within the 'design limits of power distribution is accomplished by-generating operatingilimits on the'Linear Heat Rate (LHR). and the Departure from Nucleate Boiling, (DNB). Proximity to the DNB condition is-expressed by the Departure from Nucleate. Boiling Ratio (DNBR), defined as the'ratio of the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux., The:minimum DNBR value dbring both,'normal operation ard.AOOs is the DNBR Safety Limit as calculated by the.CE-iCorrelation (Ref. 3), and corrected for such'factors as rod bow and grid spacers, and it i'sraccepted as an appropriate margin,,to DNB'for all operating *conditions. There are two'systems that~monitor'core power distribution-online:, 'the Core Operating-,Limit Supervisory System (COLSS) or-the'Core Protection.Calculators (CPCs). The COLSS and CPCs'moni'tor the core, power;diftributien'.and are capable of verifying that the LHR and,.DNBR do. not' exceed their limits. The COLSS performs this'"function. by cont-inuously monitoring I.the core power distribution and calculating core power operating limits corresponding to the'.allowable peak LHR and DNBR.,.TheCPCs perform-this, function by continuously 'calcul'ating' actual values of DNBR and local power density (LPD) 'forecomparisonwith,the respective trip setpoints. 'A DNBR penaltý factor.is 'includedJin both the COLSS and CPC -'DNBR.'cacuiation's !to -acommodate.the, effects of rod bow. 'The' amount-of:red bow in.,each assembly,-is* dependent upon the average burnup experienced by that assembly. Fuel assemblies..that-incur.higher.than average..burnup..experience greater rod bow. Conversely, fuel assemblies that receive-lower, than average burnup.experience-less rod bow-T

• 7 design calculations for a.reload core.-.-,each batchof,fuel is assigned,';a penalty that. is applied to'-the maximum integrated planar radial power':peak of theibatch;-.ýThis penalty is correlated with the,amount ofrodbow th'at is determined from the maximum average assembly burnup of the batch.

A 5single' net penal ty.for.the. COLSS and CPQ is then determined from the penalties"`as'sdciatied with each batch that comprises, a core reload, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches. (continued) PALO VERDE UNITS 1,2,3 SB:-"3.2.5-2 , I REV'_,ISI-0N. -.10

ASI B 3.2.5 BASES BACKGROUND (continued The COLSS indicates continuously to the operator howfar the .-.core is from the operatinglimits and provides an audible alarm if an operating limit is exceeded. Such a condition signifies.a reduction in the capability of the plant to withstand an anticipated~transient, but does not necessarily imply animmediate violation of fuel design limits. If the margin to fuel design limits continues to decrease, the RPS ensures that the specifiedacceptable fuel design limits are not exceeded for:AOOs by init-iating a reactor trip. TheCOLSS continually generates, an assessment of the calculated:margin for LHR and DNBRspeci-fied limits. The data required for these.assessments. include measured incore neutron flux, CEA pos.itions, and Reactor Coolant System (RCS) inlet temperature, pressure,. and flow. In addition to the monitoring performed by the COLSS, the "RPS (via the CPCs),continually infers.the core power , ',distr.ibbt.ion..and thermal margins: by,processing reactor

• coolant data.,. signals from excore neutron flux detectors,

".and.input from redundant:. reed s.witch assemblies that -.indicates CEA position'. -In this case, the CPCs assume a minimum core power of:20% RTP becaus.e-the power range excore neutron flux. detect.ing,'system is inaccurate below this power '"evel, If power d.istri:bution-orother.pa:rameters are perturbed as a 'result of. an AOO,. the high. local power densi.ty or l.ow'DNBRWtrips.:iin,,.the RP.S.initiate a reactor trip prior..to, exceeding fuel.design limits,:, The. limits onASI*F* and Tý'represent; limits within which -.the LHRand DNBR algorithmsare val:id..,, These limits are obtaneddirectly":from.the ini/tial:coreior reload analysis. I" - i." APPLICABLE The.fuel *.cl adding must not-,sustainlIdamage as a result of SAFET? ANALYSES operation,,or. AOOs (Ref:.. 4)_. ;The power distribution and CEA

i.

• erti on and alignment LC(s:preventcore power

, ".distr.ibotions from reach,ing l evel s that violate the S, *i!fboloWing fuel design criteria:

a... IDuring a. LOCA,, peak cl adding; temperature must not exceed 2200OF.(Ref.. 5);,.

c n

.*..;r

.(continued)

ý PALO VERDE UNITS 1,2,3 B 3.2.5-3 REVISION 0

ASI B 3.2.5 BASES APPLICABLE

b. During CEA misoperation events or a loss of flow accident, there must be at least SAFETY ANALYSES 95%-probability at the 95% confidence level (the 4 (continued) 95/95 DNB criterion) that the hot fuel.rod in the core does not experience.a DNB. condition (Ref..4)1;

c.

During an ejected CEA accident, the fission energy .input to the fuel must not exceed 280 cal/gm (Ref. 6): d! The control rods (excluding part strength rods) must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 7). The power density:at any, point in the core must be limited to maintain the"fuel design cniteria-(Refs. 4 and 5). This is accomhplished b*)maintaining the'poWer distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses..(Ref.-..1) with.-due regard for. the....... correlations among measured quantities, the power 'distributi6n,- and uLncertainties-in,the determination of. power distri'butioni-. ý Fuel cladding faiiur' during 'a LOCA' is l imited by restricting the maximum Linear. Heat Generation Rate (LHGR)

os that the peak cladding temperature does not exceed 2200°F (M ef:.."5)..' Pea ~c'laddiýn

+ pe at res exceeding 2200°F may .icausesevere"cladding failUre :by.oxidation due to a Zircaloy Water reactio!", The&LCOs.:governing LHRS*-ASI;,:and RCS ensure that these

criteria~ar.e.-met'i)as

'ong.'as the core is operated within the ASIand F3,, imits speiffied 'in theCOLR, and within the Tq limits% Yhe -latter.are-process variables that characterize the thr'e d'imeniofrl power distribution of the reactor ' cort&.:Operat.on withinthe limits',for these variables ensures that their actual values are within the range used in the accident analysis (Ref. 1). Fuel cladding damage does not occur from conditions outside these LCOs during normal operation.. However, fuel cladding damage results when an accident occurs due to initial conditions outside thelimits of these LCOs. This potential for fuel cladding damage exists because changes in the power distribution can cause increased power peaking and correspondingly increased local LHRs. (continued) PALO VERDE UNITS 1,2,3 -B ý3.2.5-4 REVISION 52

ASI B 3.2.5 BASES APPLICABLE SAFETY ANALYSES (continued) The ASI satisfies Criterion 2 of10 CFR 50.36 (c)(2)(ii). LCO The power distributionLCO limits are based on correlations between power peaking and certain measured variables used as inputs to LHR and DNBR operating limits. The power distribution LCO limits are, provided in the COLR. The COLR provides separate limits that are based on different combinations of COLSS and CEACs being in and out of service. The limitation on ASI ensures that the actual ASI value is maintained within the range of values used in the accident anal~yses. The ASI limits ensure that with Tq at its maximum Upper limit,,-the DNBR does not drop below the DNBR Safety Limi t for, AOOs APPLICABILITY Power dilstribution is a concern any tJime the reactor is critical. The power distr.ibution LCOs, however, are only applicable in MODE 1 above 20% RTP. The reasons these LCOs are not applicable below20% RTP are:

a.

The incore neutrin detectors that provide input to the COLSS, which then:.calculates.the.operating limits, are inaccurateodue tojthepoor,.signal to noise ratio that they experience at 6elativeli low ýore power levels. b-ýý As'a< result of this inaccuracy, the, CPCs assume a minimum,core. power. of, 20% RTP.when generating the LPD /and DNBR trip, signals.ý,.When the core power is below ,.*this level. the.core is operating well below the .,thermalI.1limi.ts andAthe resultant CPC calculated LPD and DNBR trips. are.-strongnly/,conservative. ( ontiued ""( conti nued) PALO VERDE UNITS 1-,2,3 B 3.2.5-5 ýREVISION 0

ASI B 3.2.5 BASES ACTIONS A. 1 The AS li'mits specified in the COLR-ensu rethat the-LOCA and loss'of flow 'accident: criteria assumed in the accident analyses remain valid. If the ASI exceeds its limit, a Completion Time,of 2,hours is'"allowed to,"restore the"ASI-to withfn'its. specified Iimit.- This du'ration gives the operator sufficient time' to reposition the regulating or part strength CEAs to reduce the.axial power imbalance. The magnitude'of any potential xenon oscillation is significantlyreduced if the condition is not allowed to persist for more than 2 hours'. B.1 If the ASI is not restored to within its specified limits within the required' Cbmpletiobn Time, the reactor continues. to operate with an' axial powdr distribution mismatch. 'Continued.operation i,n! this. configuration induces an axial. xenon oscillation, -,and 'resuits"-iniJncreased LHGRs when the xenon redistributes..Reducing thermal power to

• 20% RTP reduces the maximum LHR to a value that does not exceed the "fuel :design limits",'if a"d~si'gn basis event occurs.

The allowed'Completion"Timen of-4 hours is reasonable, based on operating experience., to reduce power in an orderly manner and withouit chaiiniging'plant systems. SURVEILLANCE S SR 3' 2 5*1-REQUIREMENTS The-ASI can' be monitored'by both the-incore (COLSS) and excore (CPC) neutron detector systems. The COLSS provides the operator with an~alarmi,f.,.an.ASI limit' i approached. S Veri fication 5*f the ASI:,every 12. hours ensures that the operatori s iware-of,changes,in the, ASI as they develop. A '2 '2hour"Frequency-for:,this Surveil-lance is acceptable because :themechanisms,.that. affect the ASI, such as xenon - redistribution or CEA'ddrive mechanism malfunctions, cause slow changes in the ASI,, which, can be discovered before the .limits are exceeded. (continued) PALO*VERDE UNITS 1,2,3 B 3.2.5-6 R'ýVISIOW52

RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE Design Basis Definition (continued) SAFETY ANALYSES 6, 7. Steam Generator Pressure - Low The Steam Generat6r #1 Pressure - Low and Steam Generator #2 Pressure - Low trips provide protection against an excessive rate of heat extraction from the steam generators and resulting rapid', uncontrolled cooldown of the RCS..' This trip is.needed to shut down the reactor. and-assist the ESF System in the event of an MSLB ormain.feedwater line break.accident. A main steam isolation signal (MSIS) is initiated simultaneously. 8, 9. Steam Generator Level - Low Theste/am Geiera~or #1 Level.- Low,and Steam QGenerator #2 Levl Low trips ensure that a reactor ,tr'j,.r,signal- -'is generated for the following events to .help-preientexceeding thepdesigri pressure of the RCS due-ýto the loss of -the heat sink: Inadvertent Opening of'a Steam Generator Atmospheric-Dump, Valve' (AOO).;*: 0,* Loss of. Conde6ser. Vacuum, (AOO)% .Loss of Normal-Feedwaite.r Event. (ADO):. Feedwater System Pipe Break (Accident); and, Siingle'RCPP,.Rotor, S~eizure (AOO) -11. Steam Generator Level -High9 heSteam Generator#1 Level High.,and Steam ,:'Generator #2 Level.- 'High trips are provided to protect " the2 turbine from excessive moisture carryover in case

of a steam generator overf. 11 event

..

A Main Steam Isolation -SignaT., (MS-IS):i.isAiniti-ated*: simultaneously. ,..d .(continued) PALO VERDE UNITS 1,2,3 B 3.3.1-21 REVISION 35

RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE Design Basis Definition (continued) SAFETY ANALYSES 12, 13. Reactor Coolant Flow - Low The Reactor Coolant Flow Steam Generator #1-Low and' Reactor Coolant Flow Steam Generator #2-Low trips provide protection against an RCP Sheared Shaft Event. A trip is initiated when the'pressure differential across the primary sideof either steam generator decreases below a variable setpoint. This variable setpoint stays below the pressure differential by a reset value called the step function, unless limited y a preset maximum decreasing rate determined by the Ramp Function, or a set minimum value determined by theFloor:Functi.on. The, setpoints ensure that a reactor trip occurs to limi.t fuel failure and ensure offsite doses are with"in 10 CFR 100 guidelines. .14. Local Power Density - Hi_ ýThe CPCs perform-the.cal~culations required to derive the, DNBR and LPD parameters and their associated RPS trifps. TheDNBR,- LUW and LPD High trips provide plant protection dtri'ng the following AOOs and assist the ESF systems in the mitigation of the following accidents: The LPD - High trip provides protection against fuel center'inemelttng"due to the occurrence of excessive .loca.p.wer density peaks, during the following AOOs: , 0Decrease in FeedwaterjTemperature; crease, in-.Feedwater Flow.; ..Ihtireased Main:Steam:Flow (not due to the steam line rupture): Without Turbine Trip; Uncontrolled CEA, Withdrawal From Low Power; UJncontrol.ed CEAWithdrawal at Power; and SCEA Misoperation .For the eyents listed, above (except CEA Misoperation -where,the DNBR and LPD. tripi*s.,wi 11 occur near ,,slimultar!eously)v, DNBR,. Ldw will trip the reactor first, since DNBwould. occur before fuel centerline melting would occur. (continued) PALO VERDE UNITS 1,2.3 B 3.3.1-22 REVISION 52

RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE SAFETY ANALYSES Design Basis Definition (continued)

15.

Departure from Nucleate Boiling Ratio (DNBR) - Low The CPCs perform the calculations required to derive the DNBR and LPD parameters and their associated RPS trips.. The DNBR - Low and LPD:-> High trips provide plant protection during 'the following AOOs and assist the ESF systems in the mitigation of. the following accidents.: The DNBR - Low trip provides protection against core damage due to the occurrence of~locally saturated conditions in the limiting (hot) channel during the following events and is theprimary reactor trip (trips.thel.reactor first). for these events: 0 Decrease-in Feedwater Temperature: .Y :': + 0 Increase in FeedwaterFlow; Increased Main-Steam Flow '(not due to steam line rUpture):Without:-Turbine Trip: Increased Main Steam Flow..(not due to steam line rupture),With.a Concurrent Single Failure of an . Active Component; Steam Line Break Wi.h. Concurrent Loss of Offsite ,AC Power; Loss of."s NoRmalr:AC.Powen.K; S',Partial L6ssz'of Foir.ed Reactor Coolant Flow;

• ,. " Total 'Loss -f.Forced,'Reactor Coolant Flow; Single Reacftor Coolanft Pump (RCP)

Shaft Seizure; Uncontro-l.ed CEA.:Withdrawal From Low Power; S Uhncontrolled CEA Withdrawal at Power; ,,.. CEAMi soperati on; Pri'maryy Sampleor Instrument Line Break; and Steam Generadtor Tube&.Rupture. In the. above list, bnly th. steam line break, the steamigeneratcr,-tlbe rupture, the RCP shaft seizure, and the sample or'*.ihstruineh"t::l.ine break are accidents. Th "rest are:AOOs. (continued) PALO"VERDE UNITS 1,2,3 B 3.3-1-23 REVISION.52

RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE

15.

Departure from Nucleate Boiling Ratio (DNBR)-Low SAFETY ANALYSES (continued) In the safety analysesfor transients.involving reactivity and power distributionrvanomalies, credit may be taken for the CPC VOPT auxiliary trip algorithm in lieu of the RPS VOPT trip function. The exact trip credited (CPC or RPS) is documented-in chapter 15 of the UFSAR under the individual event sections. The CPC VOPT auxiliary trip.acts, through the CPC DNBR-Low and LPD-High trip contacts to provide over power protection. When credit is taken for the CPC VOPT algorithm, the CPC VOPTlsetpoints installed in the plant are based on the,'safety analyses and may differ from the:,RPSVOPT allowable values and nominal setpoints. The setpoints associated with the CPC VOPT are controlled via Addressable Constants (TS Section -5.41) and,Reload Data Block Constants (Ref. 8 and 13). The CPC VOPTauxiliary trip al~gorithm may provide.protection-aga'inst core damage during the foil owing events: Uncibntrolle"d CEA Withdrawa'.From Low Power (AOO): Uncontrolled CEA Withdrawalbat.;Power (ADO); Single CEA Witrdrawal withini'Deadband (AOO);

• Steam Bypass Control System Misoperation (AGO):
• CEA Ejection (Accident); and

.,Main, Steam Line Break (Accident). L(continued) (conti nued) PALO VERDE UNITS 1,2,3 B 3.1-1-24, REVIS1ON:,35 I ý - ý

ESFAS Instrumentation B 3.3.5 BASES ACTIONS A. 1 and A. 2.: (continued) The Completion Time-of 1 hour allotted to restore, bypass, or trip the channel is sufficient to allow the operator to take all 'appropriate actions for the failed channel and sti.ll 'ensures that the ri-sk involved in operating with the failed.channel is acceptable. Theý failed channel must be restored to OPERABLE status prior cito'entering MODE 2 following the next MODE 5 entry. With a channel' bypassed, the coincidence logic is now in a two-dut-bf-three configuration. The Completion Time of prioý to entering MODE 2 following the next MODE 5 entry is based on adequate channel to channel independence, which allows a two-out-of-thrbe channel operation, since no single failbre willcause or prevent an ESF actuation. The intent of this requirement is that should a failure dccur that.cannot.be repaired during power operation, then continued operation i"s allowed without requiring a plant 'shutdoWn: However, the failure needs to be repaired during the next MODE 5 outage. Allowingthe unit to exit MODE 5 is acceptable, as the. appropriate retest may not be possible until'normal6perating'pressures and temperatures are -achieved. If-the*failure occurs while in MODE 5, then the problem needs to, be resolved during that shutdown, and OPERABILITY restored prior to the subsequent MODE 2.entry. Cordition Bappiiesto tOthe failure of two channels of one or more input parameters in the following ESFAS automatic trip Functions:

1.

Safety Injection Actuation Signal Containment Pressure - High Pressurizer Pressure - Low

2.

Containment Spray Actuation Signal Containment Pressure -- High High (continued) PALO'VERDE UNITS 1,2,3 B 3.3.5-23 REVISION 52

ESFAS Instrumentation .B 3.3.5 BASES ACTIONS B.1 (continued)

3.

Containment Isolation Actuation Signal Containment Pressure - High Pressurizer Pressure - Low

4.

Main Steam Isolation Signal Steam Generator #1 Pressure -. Low Steam Generator #2 Pressure - Low-Steam:Generatori#1P'Level-High Steam Generator #2 Level-High Containment Pressure-High 5, Recirculatidn Actuation Signal"," Refueling Water Storage*Tank Level - Low

6.

Auxiliary FeedWater.Actiation'Signal, SG #1 (AFAS-1) Steam Generator #1 Level - Low SG PressureýDifference.ý(SG4#2,> SG#1) - High 7.' Auxiliary FeedwatertActuation Signal SG #2 (AFAS-2) 'IISteam/ Geherator #21 Level -Low i-SG Pres'sure' Difference (SG'#1 > SG #2) - High With two inoperable channels, power operation may continue, p rovided one i'noperable 'c:hannel'is pi aced, in bypass and the other channel-is 'pladed in trip 'within 1 hour, With one channel; oft.protective, instrumentation'bypassed, the ESFAS. Functifon.i;s,i .tb-oUt-,ýf,-th'ree. logic. in'the bypassed input parameterbit' with ýanotherchahne-l failed, the ESFAS may be operafing'"with a two-out:-of-two logic::' *This is outside the assumpt idns naoe in,:the&analyses and.should be corrected. To correct the problem, the second channel is placed in tri'p..'This "psl acets "the ESFAS Function in;i'a one-out-of-two logi c.'! I.,f any of theooth'er OPERABLE channels receives a t#ip 'signal*" ESFAS actuation 1will.loccur., 'Onie of, the -two inoperable chanhels, wfll need to be restored .:toOPERABLE status', prior 'to the..next_:'required CHANNEL FUNCTIONAL TEST becaus& channel 'survei.ll:ance testing on an OPERABLE channe~l requires that-.the OPERABLE channel be

placed in b~pass!. *However,,it is not possible to bypass more than one ESFAS'channel',:'and.placinga second channel in trip will. 'result.in-an ESFAS.actuation:,Therefore, if one

.ESFAS.haneli's in trip..and asecond~channel is in bypass, a thirdinoperable channel would. place~the unit in (oLCO n u0.1. (continued) PALO VERDE UNITS 1,2,3 6 3.3.5-24 REVISION'38

.7J 1, RCS P/T Limits B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.3 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure* and temperature changes.. These loads are introduced by startup (heatup) and shutdown (cooldown) operations*,power transients, and reactor trips. This LCO limits theipressure and temperature changes during RCS-heatup and.cooldown, within the design assumptions and the~stress limits for cyclic operation. The Pressure and Temperature Limits.:Report (PTLR) contains P/T limit curves for heatup, cooldown, and inservice leak and hydrostatic (ISLH) testing, and'data for the maximum . rate of,*change of reactorcoolant temperature (Ref. 1). .Each P/T*,limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during*heatup;,or coo-ldown-maneuvering, when pressure andý'temperature-. indi'cations are monitored and .compared-to.the:yapplicable-curve to' determine that operation is within the allowable region. The LCO establishes operating limits that provide a margin -.to brittle fai.lure of-,.the reactor. vessel' and piping of the Reactor;Coolant Pressure Boundary-. (RCP.).-, The vessel is the component, most.subject, totbri1?ttle fai.lbre,, and the LCO Tlimi-ts :appl-y-,main,5l.ý to the, vessel ThZeimits do not apply .. to the -pressuri'zer,:-which.-has diff-erent, design charaCteristi.cs, and.-operati ng f-urc-ti ons,

1. CFR 50. 1Appendix,.G-(Ref. 2) -requires-,the establishment of *P/T. limits-for material f.riacture toughness requirements of the RCPB,,materia)ls,

-R.eference-2.requires an adequate margin to brittle failure during normal operation, -anticipated operati~onalo.-occurrences,.-and system hydrostatic " tests...,-1It. mandates -.the..se -of the,- ASME Code, Section III, .Appendix G. (Ref- -,3, The actual shift n the RTDT 'of the vessel material will be establ ished.-,period-cally by remov ing and,-evaluating the irradiated., reactor vessel. materi al specimens, in accordance " 'with-ASTM E-185. (Ref.. 4) and Appendix H of 10 CFR 50 S-

(Ref;y 5).:

The operating. P/T.l-i.mit curves will be adjusted, as necessary, based on the evaluation-findings and the recommendations of Reference 3. I- ,(conti nued) -PALO VERDE UNITS 1,2;3

• B 3.4.3-1 REVISION 52

RCS P/T Limits B 3.4.3 BASES BACKGROUND (continued) The P/T'limit curves are composite curves established by' superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the.most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit, Across the span of the P/T-limit curves, different-locations are more-restrictive, and, thus, the curves are composites of the most restrictive regions., The heatup curve represents a different set of restrictions than the cooldown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters-the location of the tensile stress between the outer and inner walls. The criticality limit includes the Reference 2 requirement that the limit be no. less than 40,'F above.the heatup curve or the cooldown'curve andlnot less th6n the minimum permissible'temperatureýfor inservice.l eak and hydrostatic (ISLH) testing: However, the criticality limit is not operationally limiting; a more restrictive limit exists in

• ,LC0,3.4-2,ý"RCS Minimum Temperature 'for 'Oriticality."

The consequence of vidlating the' LCO, limits is that the RCS has'be6h1operated under conditions that can result in brittl'e'faijiurre"of'theRCPB, possibly, leading to a n6nisoiabie leak:or'-ilss-ofco6lant' acc-ident. In the event -these limits are exceeded, an evaluato6n must be performed to determine the.,effect, onthe structural integrity of the RCPB compbnnt-s.' The ASME' Code, Sk'idnXI, Appendix E (Ref. 6). p6rovýieýd a recommendedsmetodo lgy for evaluating an operating' event that causes an excursion outside the limits. APPLICABLE The P/T limits are not derived'frffm Design Basis Accident SAFETY ANALYSES (OBA) Analyses. They are prescribed during normal operation 'L:, " i to av6idenhcountering press'uri. teperatire, and temperature ... rate of change cbnditions that,'might cause undetected flaws 'to propagate "ad' cause' nonductile 'failure of the RCPB, an unanalyzed condition. (continued) I. I PALO VERDE UNITS 1,2,3 B 3.4.3-2 REVISION'52

RCS P/T Limits B 3.4.3 BASES APPLICABLE Since the P/T limits are not derived from SAFETY ANALYSES any DBA, there are no acceptance limits related to the P/T (continued)" limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition. The.RCS P/T limits-satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The two elements of this LCO are:

a.

The.:limitcurves for heatup, cooldown, and ISLH testing; and, b.i.. Limits:on the rat.of change of.-temperature. The LCO limits apply to 6111 components of the RCS, except the pressurizer. These.1limi.,ts define allowable'operating regions and permit a .l arg#..nUmbor. of.op'er..ating.cycles while providing a wide. margih* to nnductile f,aiure.' -.The..limits lfor. he rate'of change:of temperature control the thermal gradient through the vessel~wall 'and are used *as inputs,.forcalculating the heatup, cooldown, and ISLH testing P/T. imit'cures... Thus,. the LCO'for the rate of change of.-*temperature restricts ~stresses Caused by thermal gradients and also en.ures thdeialidity of the P/T limit

• c u r v e s,...,

.Violati'ng the LCO limits places the reactor vessel outside of th'e.,bouhds fb the stress. an'alyses and can increase ,stres se-s in,o other RCPB'componenhts,. The-consequences depend .onseveral factors, "as foll'ows': ' a..The severity of the departure from the allowable opaing P/T 'Iegi(e' or the severity of the rate of -'. change of temperature; b., Thele'igth of.ti methe litilts.'were violated (lbnger violations allow the t*mpprature gradient in the thick vessel'. walls tosbecome monre-pronounced); and (continued) .-PALO VERDE UNITS 1,2,3 B.3.4.3-3 REVISION 0

RCS P/T Limits B 3.4.3 BASES LCO

c.

The existences, sizes, and orJentations of flaws in (continued) the vessel material. APPLICABILITY The RCS P/T limits Specification provides a definition of acceptable operation for prevention of nonductile failure in accordance with 10-CFR 50, Appendix. G (Ref. 3). Although the P/T limits, were de'velQped to provide guidance for operation during heatup or cooldown (MODES 3, 4, and 5) or ISLH testing, their Applicability is at all times, except when reactor.vessel head is fully detensioned such that the RCS cannot be pressurized,. in keeping. with the concern for nonductil:e failure.. The l.imi.ts do not apply to the pressurizer. During MODES. 1 and.:2, other. Technical Specifications provide limits for operation :.that.can,be-more restrictive than or can supplement these P/T limits. LCD03.4.1, "RCS Pressure, Temperature, and-.Flow.Departure from Nucleate Boiling (DNB) Limits ".; LCO, 3;,4.2, "RCS Mi nimumTemper'ature for Criticality:.': and Safety Limit.2.1,, -Safety Limits," also provide operational restrictions for'pressure and temperature and,maxi~mum pressure.. Furthermore, MODES 1

• and 2 are above the temperature range of..concern for

- nonduct.ilefailure,:.and stres analyses: have been performed normal maneuverj,ng; profi.es, such as power ascension or ' descent.*.... The actions of this LCO consider the premise that a ... i1atIon,*;f the im,1*its occurred durlng..normal plant ,maneuver.ing,,-'Severe :violations, caused bY abnormal tans.ients,. at-ti mes-,accompanied by..eq~uipment failures, may also requi-re addi-t.iora. actions :fromemergency operating procedureS.. (continued) I PALO VERDE UNITS 1,2,3 B 3.4.3-4 REVISION52 W

RCS P/T Limits B 3.4.3 BASES ACTIONS A.1 and A.2 Operation outside the P/T limits must be corrected so that the RCPB is returned to a Condition that has been. verified by stress analyses. The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. violations will not be severe, and the activity can be accomplishedin thisK:time in a controlled manner. Most Besides restoring operation to within limits, an evaluation is required to determine~if RCS operation can continue. The evaluation must Verify the RCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparison with pre-analyzed transients in the stre~s, analyses, new analyses, or inspection of',the'components:.. 'ASME CodeO, Section'Xi.,..Appendix E (Ref. 6), may be used support the evaluat:ioh. :However, its'use is restricted

evaluation-of the vessel beltline.

to to I ,The 72 hour ýCompletioi. T.fne is reasonabl-e to accomplish the 'evalu'ation.* The iValuation for aý,'mild.violation is possible

<within this time, but. more severeý violations may require

'special;, event speciffic stress':analyses or inspections. A favorable evaluation must be completed before continuing to operate..., ,Condition A, is 8-modified by,'a, Note: requiring Required -."Act16nh.;A.2 to be cohmpleted.,whenever thee.Condition is entered.-, The Note emphasi'zes th6"needitd perform-the

'!ýýevaluation of the effects of theeOxcursion outside the allowable limits.

Restoration.al'one~per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because:

a.

The RCS remained in an unacceptable P/T extended period of increased stress; or region for an (continued) PALO. VERDE UNITS 1,2,3 B.3.4.3-5 REVISION 52

RCSP/T Limits B 3.4.3 BASES ACTIONS B.). and B.2 (continued)

b.

A sufficiently severe event .unacceptable region. caused entry into an Either possibility,indicates a need f6r more careful examihation of the eventi'best accompl.ished with the RCS at reduced'pressure and temperature. With reduced'pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.. Pressure and temperature are MODE 3 within 6 hours and"'in < 500 ps~i a Within' 36' hours. reduced by placing the plant in MODE-5 with"-RCS pressure ' The Compl*etiorn Times are-.reasonab le-based.on operating experience, to reach the" required*'plant conditions from full power conditions in an orderly ma~hner andiWithout challenging plant systems. ' C.* 1 and C.2 2' "'The actions of' this-LCO, arn-time other th~an in MODE 1, 2. 3, or 4", cnrs.ider th& premise that a violation of the limits occurredduri'ng normal:,p.lant maneuvering.,1 Severe violations caused by abnormal transients, at times accompanied by equipment failures, may-also require additional actions',, from

• emergency operating procedures.

Operation outside the P/T limits must be corrected so that,the RCPB.is retu'rned to"a4.' condition.that has been verified by stressanalys'es.ý '.2 S..... The 'Comp1*e.tio* 'Ti" e 'immeditley reflects the urgency of restoring'-the' parameter'sk to within.. the-analyzed range. Most 'vi ol 6tions'.'wi il not' be' severe:, 'and. the* activity can be accomplished in.'a short period of time;inia controlled V (continued) PALO VERDE UNITS 1,2,3 8 3.4.3-6

.: REVISION.0

RCS P/T Limits B 3.4.3 BASES ACTIONS C.1 and C.2 (continued) Besides restoring operation to within limits, an evaluation is*:required to determine if RCS operation can continue. The evaluation must verify that the RCPB integrity remains acceptable and must be completed before continuing operation. Several methods may be used, including comparisonwith pre.-analyzed transients in the stress analyses, new:analyses,:or inspection-of the components. ASME Code, Section XI.AP'pendix E,. support the evaluation., However-evaluation of the vessel beltline'. (Ref. 6), may be used to its use is restricted to ' The, Completion Time of prior to'en'tering MODE 4 forces the evaluation prior to entering a MODE-where temperature and pressure can be significantly increased. The evaluation for afmild violation is..possible within several days, but more severe vJotions may require special, event specific stress analyses! or, inspections. Condition C is modified bj a Note 4relquing Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the "allowable limits. Restoration, alpne per.Required Action C.1 . i insufficient becaus'e h'igher.*tha'nanal'yzed stresses may -v.. have occurred, and: may tave, affected the.,RCPB integrity. SURVEILLANCE REQUIREMENTS Veri fication"nthat operation is"'within he PTLR limits is required -every.30,minute,s,wh~en.RCS pressure and temperature condfitio0ns: are prndergoning plaan.d.ch.anges. This Frequency .is::,.consi dered,reasonable iný vi~ewof, the. control room indic~ation, avaIlable-to.m6nito'K.RCS, status. Also, since temperature rate of change l imits are,speicified in hourly increments, 30 minutes permits assessment and correction for minor deviations within a reasonable time. Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied. (continued)

'PALO'VERDE UNITS 1,2,3 B. 3.4..3-7 REVISION 52

RCS P/T Limits B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 (continued) REQUIREMENTS This SR is modified by a Note that requires this SR be performed only during RCS systemrheatup,-cooldown, and ISLH testing* No SR is given for criticality-operations because LCO 3.4.2 contains a more restrictive requirement. REFERENCES

1.

TRM Appendix TA, Reactor Coolant System Pressure and Temperature Limits Report (PTLR);.,(,limits determined using methods described ;in Topical Report CE NPSD-683-A;, Revision 6, Development of a RCS Pressure and Temperature Limits Report for the Removal of P-T Limits. and LTOP:Requirements.from the Technical Specifications, April 2001). 2'! 10 CFR 50, Appendix G.

3.

ASME, Boiler and Pressure Vessel Code, Section III, Appendix.G.`

4.

ASTM E 185-82, July 1982.

5.

10 CFR 50,, Appendix H. -6.,ASME, Boiler and..Pressure..Vessel --Code,_Secti.on XI, Appendix E. PALO VERDE UNITS 1,2,3 B 3.4.3-8 REVISION"-:52

RCS Loops - MODE 4 B 3.4.6 BASES LCO Note 2 requires secondary side water temperature in each (continued) SG is < 100°F above each of the RCS cold leg temperatures .before an RCP may be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR. Satisfying the above conditionwill preclude a large pressure surge in the RCS when the RCP is started. Note 3 restricts RCP operation to no more than 2 RCPs with .RCS cold leg temperature < 200 0F, and no more than 3 RCPs with RCS cold leg temperature >200'F but *.500'F. Satisfying these conditions will.maintain the analysis -assumptionsof the flow iinduced pressure correction factors due to RCP. operation, (Ref: 1) An OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that i, s OPERABLE and has the minimum water level specified in SR 3.4.6.2. Similarly, for the SDC System,-an:,OPERABLE SDC train is composed of an OPERABLE SDC pump (LPSI) capable of providing flow to the SDC' heat;.exchanger for heat removal. RCPs and SDC pumps are OPERABLE if they are capable of being powered and are able to pro'jde. flow, if required. APPLICABILITY In MODE 4, this LCO applies because4 it is possible to remove ,core.:decay-heat.and.to provide--proper:.:boron mixing-.with:.- either the RCS loops and SGs or the SDC System. Operation in other MODES is covered by: LCO 3.4.4 "RCS Loops-MODES 1 and 2'. LCO 3.4.5, "RCS Loops - MODE 3"' (continued) PALO VERDE UNITS 1,2,3 B 3.4.6-3 REVISION 52

RCS Loops.- MODE 4 B 3.4.6 BASES APPLICABILITY LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled"; (continued) LCO 3.4,8."RCS Loops,- MODE 5, LoopsNot Filled";. LCO 31.9.4, "Shutdown Cooling (SDC) and Coolant ,Circulation -High Water Level" '(MODE 6); and LCO 3.9.5, "Shutdown Cool ing (SDC) and Coolant 'Circulation LowWater Level" (MODE 6). ACTIONS A.1 If only one required RCS loop is OPERABLE and in operation, redundancy for heat removal, is'lost. Action must be initiated immediately to r estore-a second loop to OPERABLE status. The immediate Completion Time-reflects the .importance of maintaining the availability of two paths for 4 decay heat removal.

B.1.

If only one required-SDC trainis.OPERABLE-and in operation, redundahcy for heat' removal is". dst'. "The&plant must be placed in MODE 5 within the next 24 hours. Placing the plant in MODE 5 is a conservative action.with regard to_ decay heat removal. With only one SDC trajn OPERABLE, -redundancy for decay heat removal is lost and, in the *event odf'alos's 6f'therýemainngSDC-train, it-would be safer to initiate that"oss' from MODE 5ý(* 2100F)7 rather than MODE 4 2-. O*F t.1,350 F'). 'm he Corpletfon.Timedf,24 hours is .reasonablet based Comope`atingexerieencoi to reach MODE 5 from!'MODE, 4,I! Wi th 6n1.loieSD trai~n~p~rating, in an ordr> ranead'without chaieng'fn§ plant systems. C.1 and C.2 If no RCS loops or SDC trains are OPERABLE, or in operation, all operations involving reddclo-onof RCS-boron concentration i' 'must be:..siuspended and action to restore one RCS loop.or SDC -7.train to OPERABLE status, and operation must be initiated. Boon, dilutionrequires.forced Ci'rculation for proper mixing, and the margin to criticality must"not be reduced in this type of operation. The immediate,,Completion Times reflect the importance of decay heat removal. The action to restore must .... continue until:one loop or train.-is restored to operation. (continued) PALO VERDE UNITS 1,2,3 B 3.4.6-4 .REV fLSION' 6

RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR' requires,verification every 12 hours that one trequired loop or-train is in operation and circulating reactor coolant.at a flow ratedf greater than or equal to 4000 gopri This ensures forcedflow is providing heat removal.' Verification includes flow rate, temperature, or pump status monitoring. The 12 hour Frequency has been shown by operating practice to be sufficient to regularly .assess RCS loop status. In addition, control room indication and alarms will normally indicate loop status. SR 3.4.6'.2. ThiS SR requires verification every 12 hours of secondary side water level in the required SG(s). 25% wide range. An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the reactor coolant. The 12 hour interval has been shown by operating practiceto be-suffi~c~ient, toregularly assess degradation anidiverif-y operation withijn' safety,.analyses assumptions. .'SR, 3.4.6.3( Verificat.ion "that the requi red pump is OPERABLE ensures that

an'additional RCSloop'or SDC train' can be placed in

,,ope rati*n' if ne6ded:.to maintain de8ay.,heat removal and reaotorý,,coolant circulation..Verification is performed by yerif ying properbr e ker alignmeht',and-power available to -the required.pump's.., Jhe F requehcyof 7 days is considered reasonable in view'of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES

1. PVNGS Operating"License Amendments,52, 38 and 24 for U

ni'ts 1.' 2 and;.'3,'.respec'tilely, and associated NRC SafetyEval'uati~on dated'July 25,. 1990. 2".,Not :used' 3'" PVNGS Calcu'laitibn 13-JC-SH-0200, Section 2.9. PALO-VERDE UNITS 1,2,3 ,B 3..4.6-5 REVISION 52

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• ', )
• V'

...,C',: 1.. I I I~, 'it.

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES LCO in order to use the provisions of the Note allowing the (continued) pumps to bede-energized. In this MODE, the SG(s) can be used as the backup for SDC heat removal. To ensure their availability, the RCS loop flow.path is to be maintained with subcooled liquid. In MODE 5, it is sometimes necessary to stop all RCP or SDC forced circulation., This is permitted to change operation from one SDC train to the other, perform surveillance or startup testing, perform the transition to and from the SDC, or to avoid operation below the RCP minimum net positive suction head limit. The time period is acceptable because natural circulation is acceptable for decay heat removal the reactor coolant temperature can be maintained subcooled, and boron stratification affecting reactivity control is not expected. Note.2 allows one.SDC train to be inoperable for a period of up to 2 hours provided that the other SDC train is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable train during the only time when such testing is safe and possible. Note 3 requires that secondary side water temperature in each SG is < 100°F above each of the RCS cold leg temperatures before an RCP may be started with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified in the PTLR. Satisfying the above condition will preclude a low temperature overpressure event due to a thermal transient when the RCP is started. Note 4 restricts RCP operation to no more than 2 RCPs with RCS cold leg temperature

• 200 0 F, and no more than 3 RCPS with RCS cold leg temperature > 200'F but
• 500 0 F.

Satisfying these conditions will maintain the analysis assumptions of the flow induced pressure correction factors due to RCP operation (Ref. 3). (continued) PALO VERDE UNITS 1,2,3 B3.4.7-3 REVISION 52

RCS Loops - MODE 5,. Loops Filled B 3.4.7 BASES LCO Note 5 provides-for an orderly transition from MODE 5 to (continued) MODE 4 during a planned. heatup by permitting removal of SDC." trains from operation when at least one RCP is in operation. This Note providesfor'the transition to.MODE 4 where an RCP is permitted to be in operation and replaces the RCS circulation function provided by theSDC trains. An OPERABLE SOC train-is composed-of an OPERABLE SDC pump (Cs or LPSl.) capable of providing flow to the'SDC heat exchanger for heat removal. SDC pumps areOPERABLE if they are capable of being powered and are able to provide flow (current Section XI), if required.,. A.SG can' perform as a heat sink when. itis OPERABLE and has the minimum water level specified in SR 3.4.7.2. The RCS loops may not be considered filled until two conditions needed for operation of the steam generators are met. First, the RCS must be intact. This means that all removable portions of the primary pressure boundary (e.g., manways, safety valves) are securely fastened. Nozzle dams are removed. All manual drain and vent valves are closed, and any open system penetrations (e.g., letdown, reactor head vents) are capable of remote closure from the control room. An intact primary allows the system to be pressurized as needed to achieve the subcooling margin necessary to establish natural circulation cooling. When the RCS is not intact as described, a loss of SDC flow results in blowdown of coolant through boundary openings that also could prevent adequate natural circulation between the core and steam generators. Secondly, the concentration of dissolved or otherwise entrained gases in the coolant must be limited or other controls established so that gases coming out of solution in the SG U-tubes will not adversely affect natural circulation. With these conditions met, the SGs are a functional method of RCS heat removal upon loss of the operating SDC train. The ability to feed and steam SGs at all times is not required when RCS temperature is less than 210°F because significant loss of SG inventory through boiling will not occur during time anticipated to take corrective action. The required SG level provides sufficient time to either restore the SDC train or implement a method for feeding and steaming the SGs (using non-class components if necessary). (continued) PALO VERDE UNITS 1,2,3 B 3.4.7-4 REVISION38 S

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

1. Not Used

2.

CE NPSD-770 Analysis for. Lower Mode Functional Recovery Guidelines..

3.

PVNGS Operating License Amendments 52, 38, and 24 for Units 1, 2 and 3, respectively, and associated NRC I Safety Evaluation dated"July"25, 1990.

4.

Not used. 5:. .PVNGS Calculation'-13-JC-SH-0200, Section 2.9. 1~~ 'I I ~.. I-PALO.JERDE UNITS 1,2,3 B 3.4.7-7 REVISION 52

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Pressurizer Safety Valves-MODE 4 B 3.4.11 BASES (continued) LCO One pressurizer safety valve is required to be OPERABLE in MODE 4 with no Shutdown Cooling System suction line relief valves in service. The four pressurizer safety valves are set to open 25 psia less than RCS design pressure (2475 psia) and within the ASME specified tolerance to avoid exceeding the maximum RCS design pressure SL to maintain accident analysis assumptions, and to comply with ASME Code requirements. The limit protected by this specification is the Reactor Coolant Pressure Boundary (RCPB) SL of 110% of design pressure. Inoperability of all valves could result in exceeding the SL if a transient were to occur. The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components. increased leakage, or additional stress analysis being required prior to resumption of reactor operation. APPLICABILITY In MODE 4 abovethe LTOP System'temperatures OPERABILITY of one valve is required. MODE 4 s conservatively included, although the listed accidents may not require a safety valve for protection. The requirements for overpressure protection in other MODES and in MODE 4 at or below the LTOP System temperatures are covered by LCOs 3.4.10, "Pressurizer Safety Valves - MODES 1, 2 and 3," and LCO 3.4.13, LTOP System. The Note allows entry into MODES 3 and 4 with the lift settings outside the LCO limits. This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 72 hour exception is based on 18 hour outage time for each of the four valves. The 18 hour period is derived from operating experience that hot testing can be performed within this timeframe. (continued) PALO VERDE UNITS 1,2,3 B 3.4.11-3 REVISION 0

Pressurizer Safety Valves-MODE 4 B3.4.11 BASES (continued) ACTIONS A.1, A.2, and A.3 If all pressurizer safety valves are inoperable, the plant must be, brought to a condition.where overpressure protection is.provided, then to a MODE, in.whic~hthe requirement does not apply..*, To achieve this"status, one Shutdown Cooling System suction line relief must.be placed in service immediately, then the plant must be brought to at least MODE 4 with any RCS cold leg temperature less than or equal to the;LTOP enable temperature specified in the PTLR within 8 hour's so thatLCO 3.4.13 (LTOP System) would apply. It is reasonabl.e to pursuethe, ACTION to.place a shutdown 2cooling.system suction relief.Valve in service immediately (without delay) because, the plant is already within the shutdown cooling system entry temperature.of less than 350'F.- The Completion -Time'6f immediatelY requires that the required action be pursued without delay, and in a controlled

manner, and reflects the impor'tanceof maintaining the RCS overprotection system,.

The,8 hours allowed to be in MODE 4 with any.RCS temperatureliess thaýn o'r.eqcfal to the LTOP enable,-temperature' specified i.n thePTLR'is reasonable, basedion,.operaoing,.experienice, td 'rec'h 'this condition without"'-challenging plan~t systems. For',the-Shuitdown Cool,ingSystem suction:ine relief valve -that is;?equir'ed to be in service in accordance with Required Action Al:, SR 3.4.11.2 and SR'3:4.11.3 must be performed or verfi&edpefformmed'withihn 12'hours. This ensures thatthe required Shutdown-Cooling System suction line relief valva! is OPERABLE. A ShutdoWh Cooling System sucbioK, line relief 'vahlve is: OPERABI.Eýwhe~n its isolation y ave. oreiopen,, itslift Setpoint. is,' set at 467 psig or 1, less, 7and'*tsting 'has!proven J ts'. abJlJ. t"to open at that setpoint. ` r t a__! o o If the Required Actions and associated Completion Times are not met, overpressurization is possible. The 8 hours Completion Time to be in MODE 4 with any RCS cold leg temperature less than or equal to the LTOP enable temperature specified-in the PTLR places the unit in a condition where the LCO does not apply. '(contihned) PALO VERDE UNITS 1.,2,3 B 3.'4.11-4 REVISION'ý2

• :.l~-~

? LTOP System B 3.4.13 BASES BACKGROUND (continued) Shutdown Cooling System Suction.Line Relief Valve. Requirements (continued), When a Shutdown Cooling System suction-1ine relief valve lifts due to an increasing pressure transient, the release of coolant causes the pressure increase to slow and reverse. As the Shutdown Cooling System suction li-ne relief valve releases-coolant, the'system-pressure decreases until valve reseat pressure 'is readhed and the Shutdown Cooling system suctifn line relief valve cl6ses'. i* At low teMperatOres with the Shutdown Cooling System *suction line relief valves aligned to the RCS, it is necessary to

restrict heatup~and cooldown rates toassure that P-T limits are not exceede:d These P-T limits are usually applicable to'a finite time period suchas on*e ccle -5 EFPY, etc. and are based upon irradiation damage prediction by the end of the period.

Accordingly, eaih time P-T*.limits change, the eTOPcSysteam.ey, d"to'be reanalyzed andtmodified, if necI essary, nton be 0*.its fdunctidon'. ie. On.' ce the RCS i-s depressurized', a vent exposed to the .containment atmosphere wil... I maintain the RCS at containment ambient-pressure in anRCS-*overpressure transient, if the relieVing 'requirements'of the :tran~ient do not exceed the capabilities of-the Vent. Thus; the"vent path must be ..,capable of relieving the flow resulting from the limiting. L*OP mass or heat input transient"andm'aintaining pressure below 'the PITlim'ts-The 'requ'ired-vent capacity may be provided by, one. or more-vent'paths. ior. n .RCSlen tdflO capacity, it irequies removing'al.pres'surizerisafety valves, or .sMlarly stabl ishinig a" v-nt v by opnihg bthe pressurizer maway, (Rej. -11),. The vent.'path.<sD must'be above the level of reactor coolagnt. so 'as hOt to drainzthe RCS when open. (continued) PALO.VERDE UNITS 1,2i3 B 3.4.13-3 PEVISION.1

LTOP System B 3.4.13 BASES APPLICABLE SAFETY ANALYSES Safety analyses (Ref. 3).,demonstrate that the reactor vessel is adequately protected against exceeding the Reference 1 P/T limits during shutdown. In MODES 1, 2, and 3,. and in MODE 4 with any RCS cold leg temperature greater than the LTOP enable temperature specified in the PTLR, the pressurizer safety valves prevent RCS pressure from exceeding the Reference.1 :limits. -At the LTOP enable temperature specified in the PTLR and below, overpressure prevention falls to the OPERABLE Shutdown Cooling System suction line relief Valves or. to adepressurized RCS and a sufficient sized RCS vent. Each of these means has a limited overpressure' relief capability. II The actual temperature at which the;pressure in the P/T .limit, curve falls below the pressurizer~safety valve setpoint increases as the reactor vessel material toughness decreases due toneutron embrittlement. Each time the P/T -limit curves are revised. the.LTOP System will be re-evaluated to ensure, itsfunctional requirements can still be satisfied using the Shutdown.Cooling System suction line relief valve method or-the depressurized,-and vented RCS condition. Reference. 3 contains,,the acceptance-limits that satisfy the LTOP requirements. Any change tothe RCS must be evaluated against these analyses to determine the impact of the change on.;: the LTOP. acceptance, l:imits-. Transients that are capable ofoverpressurizing the RCS are, tf.atego.rized as:, el+her mass or heat :input transients, examples-of which fol Iow: Mass InpUt: Type Transients,,

a.

Inadvertent safety injection: or , b. i Charging/letdown flow mismatch.,. Heat InputTypeTransients.. a.. ,Inadvertent actuationof-pressurizer heaters; b... Loss of'shutdown cooling (SDC.); orý. .. *c. Reactor.coo:lant) pump (R.CP) startupkqwith temperature asymmetry within the RCS-,or between the RCS and steam generators, (continued) PALO VERDE UNITS 1,2,3 B 3.4.13-4 REVISIOW52

7:'q q; LTOP System BASES B 3.4.13 APPLICABLE SAFETY ANALYSES (continued) f References 3, 7, 8 and 9 analyses demonstrate that either one Shutdown Cooling System suction line relief valve or the RCS vent can maintain RCS pressure below limits for the two most limiting analyzed events:

a.

The start of an idle RCP with secondary water temperature of the SG

• 1000 F above RCS cold leg temperatures.

,-b. An inadvertent SIAS with 'a water solid RCS, three letdown isolated. two HPSI pumps injecting into charging pumps injecting, and Fracture mechani.cs.analyses-established the temperature of LTOP Applicability*atlessthan or equal to the LTOP enable temperature specified in.the PTLR.

Above these temperatures, the pressurizer safety valves provide the reactorK'vessel.pressure protection., The vessel materials "were'assumed'tohave a neutron irradiation accumulation equal'to-the effective full power years of operation specif.ied'In the PTLR.

The consequences of a small break Loss Of Coolant Accident '(LOCA) in LTOP MODE 4-conform to.10 GFR 50.46 and 10 CFR 50. Appendix. K (Refs.: 4.'and. 5.). The fracture mechanicsanalyses'show..that the vessel is protected when the Shutdown Cooling System suction line 'relief valves are:set to open" at or below 467 psig. The setpoint.iS deri-ved bylrmodeljing:.the,*performance of the LTOP System, assuming, the limiting al.owed'*LTOP transient. The Shutdown Cooling System suction line relief valves setpoints at or below the.*den1'i~ved.,li~mit* ensure *the, Reference 1 limits will be met. The Shutdown Cooling System suction line relief valves setpoi'nts'wi'l',be re-evaluated-.for compliance when the revised P/T limits conflict with the LTOP analysis limits. The P/T limits are.,perioldica.l yK.modifi.ed. as the reactor vessel material toughness decreases due to embrittlement caused by neutron.irradiati'on:..,VRevised.,P/T limits are determined using neutron fluence projections and the results of'examinationsof thei.reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3, "RCS Pressure and.Temperature-*(P/T)'Limits.",discuss these (continued) PALO' VERDE UNITS 1,2,3 B 3.4.13-5 REVISION 52

LTOP System B 3.4.13 BASES APPLICABLE The Shutdown Cooling System suction line relief valves are SAFETY ANALYSES considered active components. Thus, the failure of one-(continued) Shutdown Cooling System suction line relief valve represents the worst case, single active failure. RCS Vent;-Performance-With the RCS depressurized, analy'ses show:a vent size of 16 square inches is capable of mitigating the limiting allowed LTOP overpressure transient. In that event, this size vent maintains RCS pressure less than the maximum RCS pressure on the P/T limit curve. The RCS vent size will also be re-evaluated for compliance each time the PiT limit curves are revised based on the results of the vessel material surveillance. The RCS vent is passive and is not subject to active failure. LTOP System satisfies Criterion 2 of 10.CFR 50.36 LCO This LCO is required to ensure:that--the LTOP System is OPERABLE. The LTOP System is. OPERABLE when the pressure relief capabilit:ies are OPERABLE. Violation of this LCO could lead to the loss of low temperature overpressure mitigati-6n and violatiOn'*of the Referencel1 limits as a resultdof an operattonal: transient. Theelementsof,the LCO that provide overpressure mitigation through pressure reiief 6,re: a, Two OPERABLE Shutdown Cooling System suction line relief valves; or

b.

The depressurized RCS and an RCS vent. A Shutdown Cooling System suction line relief valve is OPERABLE for LTOP when its isolation valves are open, its lift setpoint is set at 467 psig or less and testing has proven its ability to open at that setpoint. An RCS vent is OPERABLE when open with an area ! 16 square inches. For an RCS vent to meet the specified flow capacity, it requires removing all pressurizer safety valves, or similarly establishing a vent by opening.the pressurizer manway (Ref. 11). The vent path(s) must be above the level of reactor coolant, so as not to drain the RCS when open. (continued) PALO VERDE UNITS 1,2.3 B 3.4.13-6 REVISION 0

LTOP System B 3.4.13 BASES LCO Each of 'these methods of overpressure prevention is capable (continued) of mitigating the limiting LTOP transient. The Note requires that, before an RCP may be started, the secondary side water temperature (saturation temperature corresponding to SG pressure) in each SG is

• 100°F above each of the RCS'cold leg temperatures... Satisfying this SonditiOn will preclude a large pressure surge in the RCS when the RCP is started.:-

APPLICABILITY This LCO is applicable in MODE 4 when, the temperature of any RCS cold leg is less than or equal to the LTOP enable temperature specified in thfe PTLR, in MODE 5, and in MODE 6 whenithe, reactor vessel headAs on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 P/T ;limits.above the LTOP enable temperature. The requirements for overpressure protection in MODES 1, 2 .and 3, and in MODE 4 above the LTOP System temperatures are covered by. LCO 3.4.10-"Pressurizer Safety Valves MODES 1, 2, and 3," and LCO 3.4.11,"Pressurizer'Safety Valves MODE 4." Whernthe' reactor vessel head is off overpressuri~zat.i,on cannotoccur. I LCO'3.4.3 provides the operational P/T'l*i.mits for all MODES Low temperature overpressure preventti-on is most critical during shutdown when the RCS.,s water solid,. and a mass or heat input transient can. cause a very 'rapid increase in RCS pressure wh'en little orno -timeallows.oerator action to mitigate the event. .2 I I' (continued) PALO VERDE UNITS 1,2,3 .B,3.4.13-7 REVISION 52

LTOP System B 3.4.13 BASES ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable LTOP system. There is an increased risk associated with entering MODE 4 from MODE 5 with LTOP inoperable:.,and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of the risk assessment addressing inoperable the systems and components, should not be applied in this circumstance. A.1 In MODE 4 when any RCS Cold leg temperature is less than or equal to the LTOP enable temperature specified in the PTLR. with one Shutdown Cooling System suction line relief valve inoperable, two Shutdown',Cooli.ng System suction line relief valves must berestored toOPERABLE. status within a " Completion Time of,7 day;s'.g Two val.ves are required to meet the LCO requirement and to provide low temperature overpressure mitigation while withstanding a single failure of an active component. The Completion Time is based,on the facts that only one Shutdown Cooling Systemlsuction line relief valve is required to mitigate'an overpressure transient and that the likelihood-of an active failure of the-remaining valve.path during this time period is very low. B.1 The consequences of operational events that will overpressure the RCS are more severe at lower temperature (Ref. 6). Thus, one required Shutdown Cooling System suction line relief valve inoperable in MODE 5 or in MODE 6 with the head on, the Completion Time to restore inoperable valve to OPERABLE status is 24 hours. The 24 hour Completion Time to restore two Shutdown Cooling System suction line relief valves OPERABLE in MODE 5 or in MODE 6 when the vessel head is on is a reasonable amount of time to investigate and repair several types of Shutdown (continued) PALO VERDE UNITS 1,2,3 B 3.4.13-8 REVISION".,52

,,-{ C* LTOP System BASES. B 3.4.13 REFERENCES (continued) 8. Pressure Transient Analyses

a. V-PSAC-009 (3876 MWt w/Originali Steam Generators)

b. MN725-00118 (Unit 2, 4070 MWt w/Replacement Steam

.,Generators)

c. MN725,00562'(Units 31, 4070 MWt w/Replacement Steam Generators)

9.

Mass Input Pressure-Transient in Water Solid RCS

a. V-PSAC-010 (387.6 MWt w/Original:Steam Generators) b.; MN725-001-17 (Unit 2, 4070 MWt w/Replacement Steam

,-.Generators) c,. MN7,25-01495:(Units 3 1-4070 MWt w/Replacement Steam 10..M, BGeneratorsn ) P s VsSc n

10.

ASME, Boiler and Pressure Vessel. Code, Section XI.

11, 13-ýC0O-93_016. ,Pa thsvs'. Days Sensiti'vity Study 'on Pressurizer Vent Post,ShutdoWn. PALO VERDE UNITS 1,2,3 '-B 3.4.13-11 REVISION 52

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'1 CREATCS B 3.7.12 BASES (continued) ACTIONS B.1 and B.2 (continued) In MODE 1, 2, 3, or 4, when-Required Action A.1 cannot be completed within the required Completion Time, the unit must be placed in a MODE that minimizes the accident risk. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. C.1 In MODE 5 or 6, if Required Action A.1 cannot be completed within the required Completion Time, the OPERABLE CREATCS train must be placed in operation immediately. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that.any active failure will be readily detected. D.1 and D.2 During movemernt of irradiated fueliassemblies, if Required Action A.1 cannot be completed within the Required Completion Time, the OPERABLE CREATCS train must be placed in operation immediately or movement of irradiated fuel assemblies must be suspended immediately. The first action ensures that the remaining train is OPERABLE, that no undetected failures preventing system operation will occur, and that any active failure will be readily detected. If the system is not immediately placed in operation, this action requires suspension of the movement of irradiated fuel assemblies in order to minimize the risk of a release of radioactivity that might require isolation of the control room. This does not preclude the movement of fuel to a safe position. E.1 and E.2 In MODE 5 or 6, or during movement of irradiated fuel assemblies with two CREATCS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes the accident risk.- This does not preclude the movement of fuel to a safe position. (continued) PALO VERDE UNITS 1,2,3 B 3.7.12-3 REVISION 52

CREATCS B 3.7.12 BASES (continued) ACTIONS F.1 (continued). If both CREATCS trains are inoperable in-MODE 1. 2, 3, or 4, the CREATCS mnay not be capable of performing the intended function and the unit is.in a condition outside theaccident, analysis., Therefdre, LCO 3.0.3 must~be entered immediately SURVE.ILLANCE SR -.3.,7.12.1 REQUIREMENTS. This SR veri'fies that the heat removal capability of the system, is sufficient to meet design requirements. This SR consists of, a combinationoQf testingandý,calculations. An 18 month Frequency is, appropriate, sincesignificant degradation ofthe CREATCS,,is slow, and is not expected over this time period. REFERENCES-1 UFSAR. Section' 9.'4. PALO VERDE UNITS 1,2,3 B 3.7.12-4 REVIS.IONi.10,

Spent Fuel Assembly Storage B 3.7.17 B 3.7 PLANT SYSTEMS B 3.7.17 Spent Fuel Assembly Storage BASES. BACKGROUND The spent fuel storage is designed to store either new (nonirradiated) nuclear fuel assemblies, or burned

• (irradiated)'fuel assemblies in a vertical.configuration..

underwater, The storage pool was originally designed to store up to 1329 fuel assemblies in a borated'fuel storage mode.' .,The current storage configuration, which allows credit to be taken for boron'concentration, burnup, and decay time, and "does not require-neutron absorbing (boraflex) storage cans, provides for a maximum storage of 1209 fuel assemblies in a four-region donfiguratjon.".The. design basis of the spent fuel cooling system, however, is'to provide adequate cooling to the spent fuel during all operating conditions (including full core offload) for only 1205 fuel assemblies (UFSAR section 9.1.3)T, Therefore-, an additional four spaces are mechanically. blocked~to limit the maximum number of fuel assemblies that may be stored in the spent fuel storage pool to 1205. Region 1 is comprised of two 9x8 storage racks and one 12x8 storage rack. Cell blocking devices are placed in every other storage cell location in Region 1 to maintain a two-out-of-four checkerboard configuration. These cell blocking devices prevent inadvertent insertion of a fuel assembly into a cell that is not allowed to contain a fuel assembly. Region 3 is comprised of three 9x8 storage racks and one 9x9 storage rack in Units 2 and 3. Region 3 is comprised of five 9x8 storage racks and one 9x8 storage rack in Unit 1. Since fuel assemblies may be. stored in every Region 3 cell location, no cell blocking devices are installed in Region 3. Regions 2 and 4 are mixed and are comprised of seven 9x8 storage racks and three 12x8 storage racks in Units 2 and 3, Regions 2 and 4 are mixed and are comprised of five 9x8 storage racks and three 12x8 storage racks in Unit 1. Regions 2 and 4 are mixed in a repeating 3x4 storage pattern in which two-out-of-twelve cell locations are designated Region 2 and ten-out-of-twelve cell locations are designated Region 4 (see UFSAR Figures 9.1-7 and 9.1-7A). Since fuel assemblies may be stored in every Region 2 and Region 4 cell location, no cell blocking devices are installed in Region 2 and Region 4. (continued) PALO* VERDE UNITS 1,2,3 B.3.7.17-1 REVISION. 52

Spent FuelAssembly Storage B 3.7.17 BASES BACKGROUND The spent fuel storage cells, are installed in.parallel rows (continued) with a nominal center-to-center. spacing'of 915 inches. This spacing, a minimum soluble boron concentration of 900 ppm, and*the storage of fuel in the appropriate region based on assembly burnup in accordance with TS Figures 3.7.17-.1, 3.7.17-2, and 3.7.17-3 is sufficient to maintain a kef'iOf'

• ,0.95 for fuel of original'maximum radially averaged enrichment ofup to 4.80%.

APPLICABLE: ' The-spent fuel storage pool is designed for non-SAFETYANALYSES criticality.by use of adequ.ate spacing, credit for boron cdocentrationa.6nd the storage-of fuel.in the appropriate region based ornass~mbly bornup in'accordance with TS Figures 3.7.17-1, 317.17-2,.and 3.,717-3. The design requirements related to,criticality (TS..4.3.1.1) are keff<*1l.10 as)sumingqno'credit-for'boron and kerf

• 0.95 taking credit for-soluble bor6n.- The burnup versus enrichment requi'rementsi_(TS:'Figures-3:7.17-1, 3.7,17-2, and 3.,7.17-3) are developed. assuming keff<'1.0 with no credit taken for solublebbron,' and that:keff{- 0.95 assuming a soluble'boron Con'6entration"of 900' pm'and the most limiting single fuel mishandling accident..

The analysis of the reactivity effects of fuel storage in_ the spent fuel storage racks was performed by ABB-Combustion EngIneering-.(CE),using the three-dimensional Monte-Carl'o'c6de KENO'-VA-wi;th the updatedI44 groupn ENDF/B.-5 neutron cross section library.,-'The KENO 'ode' ha6 bee'previously used by CEforth'eai'alysi.s aouel rackeactivity and have been benchmarked against' results from niumerods critical experiments,.- These experiments simulate-the PVNGS fuel , storage racks 'as realistical y, as'possible with respect to , paramet:rs important t6 rac-ti'vity'-&ih'aS enrichment and assembl-y',spacing.,) T.he tmod6li,g 'b6f.,Regi..ons 2, 3,, and.4 included several qo~eratve~ssmpi6n,.. se,:ssumpýýions neglected the c~onservative. ass~ump sh~' tyieff6tsý of. poison'shims in' the assemblies and reattiv. pinefheefssmtin egetech structural grids.. T-hese assumptions tehd to increase the cal cu ated :iffective mu~itipl. icati on '-adtor (keff) of the racks. The stored fue"'assemblies were'modeled as CE 16x16 assemblies with a nominal pitch of 0.5065 inches between fuel rods, a fuel pellet diameter of 0.3255 inches, and a UO(2) density of 10.31 g/cc. (continued) II PALO VERDE UNITS 1,2,3 B 3.7.17-2 REVISION:3

~1' Spent Fuel Assembly Storage B 3.7.17 BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent-fuel pool. ACTIONS. A.1 Required Action A.I1 is'modified by a'Note indicating that LCO 3.0.3 does not apply. When the configuratiOn of fuel assemblies stored in-the-spent fuel pool is not in accordance with Figures 3.7.17-1, 3.7,1-2, and..3:7.17,3, immediate action must be taken to*.. make the necessalry fuel,.assembly.movement(s) to bring the

• configuration i "to compliancei.with Figures 3.7.17-1,

' 3.7..17-2 a "d -ý,3. 17 3. I If moving-irradiated fuel assemblies.While in MODE 5 or 6, ....LCO 38,0i3 wouldnot specify any acti.on.. If moving irradiated fuelý-assemblies whie. in. MODE!1.,.3, 'r 4, the fuel 'mvmetiindependeno f!i:

2. 3per4,tefl

,movementher is nabltyoreactr operation. Therefore, in eithe i 1ai t ovy e fuel assemblies is not sufficient reason to requi.re areactor. shutdown. ,.SURVEITLLANCE; ,, REQUIIREMENT1S' SR 3,7.17.1

I ThiS

' TR e&iis-,by..ad ini.st r -t i ve, means~ that enri.chm~ent,,and 'bunu of the 'fuj` Iassibly yis With1 Tiue3~4- '.K~2

• and37,17-3

acc,

-onTnyn LCfi Sei ioni~ 4.ý3.1..1. the initial in accordance in the To Ima'nialIy,de~ter-mine-t:he 6l l.Qwed. SFP"r'egion for a fuel ass.61biy the actu;al6 urn'up -1.iss"-oypar1d -to the burnup " requiuremdnt Ifor' the gi en' ii6itia l',.ehiri'chment and appropriate decay time from Figur'63:.:17-1, 3.7.17-2, or 3.. ... 17-3. if the actual burnup is grea.ter than or equal to the' burnup requirem ent 'then t .he "fuel a:'ssembly is eligible t.

• be 'stQred in the,corftspqonding regio on. If the actual brnup is less than.the burnup.equirement, *then the comparison.needs'to be repeated uising 'another curve for a lower' number*ed region.

Note the following: (continued) PALO VERDE. UNITS 1,2,3 B 3..7,17-.5 REVISION 3

,~ Spent Fuel Assembly Storage B 3.7.17 BASES SURVEILLANCE that a fuel assembly that does not meet the burnup REQUIREMENTS requirement for Region 2 must be stored in Region 1, (continued)

• that any fuel assembly may be stored in Region 1,
• that any fuel assembly may be stored in a lower numbered region than the region for which it qualifies because burnup requirements decrease as region numbers decrease (refer also to Tech Spec 4.3.1.1),

and that comparing actual burnup to the burnup requirement for zero decay time will always be correct or conservative. REFERENCES

1.

UFSAR, Sections 9.1.2 and 9.1.3.

2.

PVNGS Operating License Amendments 82, 69, and 54 for Units 1, 2, and 3 respectively, and associated NRC Safety Evaluation, dated September 30, 1994.

3.

Letter to T. E. Collins, U.S. NRC to T. Greene, WOG, "Acceptance for Referencing of Licensing Topical Report WCAP-14416-P, Westinghouse Spent Fuel Rack Methodology (TAC NO. M93254)", October 25, 1996.

4.

13-N-001-1900-1221-1. "Palo Verde Spent Fuel Pool Criticality Analysis," ABB calculation A-PV-FE-0106, revision 03, dated January 15, 1999.

5.

Westinghouse letter NF-APS-10-19, "Criticality Safety Evaluation of the Spent Fuel Pool Map with a Proposed Region 3 Increase," dated.February 25, 2010. PALO VERDE UNITS 1,2,3 B 3.7.17-6 REVISION 52}}